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Sample records for german containment code

  1. COCOSYS: Status of development and validation of the German containment code system

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Arndt, S.; Klein-Hessling, W.; Schwarz, S.; Spengler, C.; Weber, G.

    2006-01-01

    For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of a severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants. COCOSYS is composed for three main modules, which are separate executable files. The communication is realized via PVM (parallel virtual machine). The thermal hydraulic main module (THY) contains several specific models relevant for the simulation of severe accidents. Beside the usual capabilities to calculate the gas distribution and thermal behavior inside the containment, there are special models for the simulation of Hydrogen deflagration, pressure suppression systems etc. Further detailed models exist for the simulation of safety systems, like catalytic recombiners (PAR's), safety relief valves (used in WWR-440/V-230 type plants), ice condenser model, pump and spray system models for the complete simulation of cooling systems. The aerosol and fission product part (AFP) describes the aerosol behavior of nonsoluble and as well as hygroscopic aerosols, iodine chemistry and fission transport. Further the decay process of nuclides is considered using ORIGIN like routines. The corium concrete interaction (CCI) main module is based on an improved version of WECHSL extended by the ChemApp module for the

  2. Containment Code Validation Matrix

    International Nuclear Information System (INIS)

    Chin, Yu-Shan; Mathew, P.M.; Glowa, Glenn; Dickson, Ray; Liang, Zhe; Leitch, Brian; Barber, Duncan; Vasic, Aleks; Bentaib, Ahmed; Journeau, Christophe; Malet, Jeanne; Studer, Etienne; Meynet, Nicolas; Piluso, Pascal; Gelain, Thomas; Michielsen, Nathalie; Peillon, Samuel; Porcheron, Emmanuel; Albiol, Thierry; Clement, Bernard; Sonnenkalb, Martin; Klein-Hessling, Walter; Arndt, Siegfried; Weber, Gunter; Yanez, Jorge; Kotchourko, Alexei; Kuznetsov, Mike; Sangiorgi, Marco; Fontanet, Joan; Herranz, Luis; Garcia De La Rua, Carmen; Santiago, Aleza Enciso; Andreani, Michele; Paladino, Domenico; Dreier, Joerg; Lee, Richard; Amri, Abdallah

    2014-01-01

    The Committee on the Safety of Nuclear Installations (CSNI) formed the CCVM (Containment Code Validation Matrix) task group in 2002. The objective of this group was to define a basic set of available experiments for code validation, covering the range of containment (ex-vessel) phenomena expected in the course of light and heavy water reactor design basis accidents and beyond design basis accidents/severe accidents. It was to consider phenomena relevant to pressurised heavy water reactor (PHWR), pressurised water reactor (PWR) and boiling water reactor (BWR) designs of Western origin as well as of Eastern European VVER types. This work would complement the two existing CSNI validation matrices for thermal hydraulic code validation (NEA/CSNI/R(1993)14) and In-vessel core degradation (NEA/CSNI/R(2001)21). The report initially provides a brief overview of the main features of a PWR, BWR, CANDU and VVER reactors. It also provides an overview of the ex-vessel corium retention (core catcher). It then provides a general overview of the accident progression for light water and heavy water reactors. The main focus is to capture most of the phenomena and safety systems employed in these reactor types and to highlight the differences. This CCVM contains a description of 127 phenomena, broken down into 6 categories: - Containment Thermal-hydraulics Phenomena; - Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena; - Aerosol and Fission Product Behaviour Phenomena; - Iodine Chemistry Phenomena; - Core Melt Distribution and Behaviour in Containment Phenomena; - Systems Phenomena. A synopsis is provided for each phenomenon, including a description, references for further information, significance for DBA and SA/BDBA and a list of experiments that may be used for code validation. The report identified 213 experiments, broken down into the same six categories (as done for the phenomena). An experiment synopsis is provided for each test. Along with a test description

  3. APR1400 Containment Simulation with CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Chung, Bub Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  4. APR1400 Containment Simulation with CONTAIN code

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Chung, Bub Dong

    2010-01-01

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  5. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1983-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  6. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1982-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  7. CONTAIN code analyses of direct containment heating experiments

    International Nuclear Information System (INIS)

    Williams, D.C.; Griffith, R.O.; Tadios, E.L.; Washington, K.E.

    1995-01-01

    In some nuclear reactor core-melt accidents, a potential exists for molten core-debris to be dispersed into the containment under high pressure. Resulting energy transfer to the containment atmosphere can pressurize the containment. This process, known as direct containment heating (DCH), has been the subject of extensive experimental and analytical programs sponsored by the U.S. Nuclear Regulatory Commission (NRC). The DCH modeling has been an important focus for the development of the CONTAIN code. Results of a detailed independent peer review of the CONTAIN code were published recently. This paper summarizes work performed in support of the peer review in which the CONTAIN code was applied to analyze DCH experiments. Goals of this work were comparison of calculated and experimental results, CONTAIN DCH model assessment, and development of guidance for code users, including development of a standardized input prescription for DCH analysis

  8. Hydrogen burn assessment with the CONTAIN code

    International Nuclear Information System (INIS)

    van Rij, H.M.

    1986-01-01

    The CONTAIN computer code was developed at Sandia National Laboratories, under contract to the US Nuclear Regulatory Commission (NRC). The code is intended for calculations of containment loads during severe accidents and for prediction of the radioactive source term in the event that the containment leaks or fails. A strong point of the CONTAIN code is the continuous interaction of the thermal-hydraulics phenomena, aerosol behavior and fission product behavior. The CONTAIN code can be used for Light Water Reactors as well as Liquid Metal Reactors. In order to evaluate the CONTAIN code on its merits, comparisons between the code and experiments must be made. In this paper, CONTAIN calculations for the hydrogen burn experiments, carried out at the Nevada Test Site (NTS), are presented and compared with the experimental data. In the Large-Scale Hydrogen Combustion Facility at the NTS, 21 tests have been carried out. These tests were sponsored by the NRC and the Electric Power Research Institute (EPRI). The tests, carried out by EG and G, were performed in a spherical vessel 16 m in diameter with a design pressure of 700 kPa, substantially higher than that of most commercial nuclear containment buildings

  9. Integrated severe accident containment analysis with the CONTAIN computer code

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

    1985-12-01

    Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

  10. Direct containment heating models in the CONTAIN code

    International Nuclear Information System (INIS)

    Washington, K.E.; Williams, D.C.

    1995-08-01

    The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale

  11. Direct containment heating models in the CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Washington, K.E.; Williams, D.C.

    1995-08-01

    The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.

  12. Hydrogen management techniques in German LWR-containments

    International Nuclear Information System (INIS)

    Berg, H.P.; Froehmel, T.

    1993-01-01

    Investigations are described which are necessary to develop an accident management concept for German PWRs, in particular possible solutions of the hydrogen problem resulting from a core melting accident. This work is an important part of the Nuclear Regulatory Research Programme initiated and financed by the Federal Office for Radiation Protection (BfS). Two fundamental strategies are discussed: prevention of the formation of inflammable gas mixtures by making the atmosphere of the containment inert, and mitigation of the consequence of possible combustion by limiting the local hydrogen concentration. (Z.S.) 1 fig

  13. 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

  14. CONTEMPT-DG containment analysis code

    International Nuclear Information System (INIS)

    Deem, R.E.; Rousseau, K.

    1982-01-01

    The assessment of hydrogen burning in a containment building during a degraded core event requires a knowledge of various system responses. These system responses (i.e. heat sinks, fan cooler units, sprays, etc.) can have a marked effect on the overall containment integrity results during a hydrogen burn. In an attempt to properly handle the various system responses and still retain the capability to perform sensitivity analysis on various parameters, the CONTEMPT-DG computer code was developed. This paper will address the historical development of the code, its various features, and the rationale for its development. Comparisons between results from the CONTEMPT-DG analyses and results from similar MARCH analyses will also be given

  15. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  16. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  17. DCH analyses using the CONTAIN code

    International Nuclear Information System (INIS)

    Hong, Sung Wan; Kim, Hee Dong

    1996-08-01

    This report describes CONTAIN analyses performed during participation in the project of 'DCH issue resolution for ice condenser plants' which is sponsored by NRC at SNL. Even though the calculations were performed for the Ice Condenser plant, CONTAIN code has been used for analyses of many phenomena in the PWR containment and the DCH module can be commonly applied to any plant types. The present ice condenser issue resolution effort intended to provide guidance as to what might be needed to resolve DCH for ice condenser plants. It includes both a screening analysis and a scoping study if the screening analysis cannot provide an complete resolution. The followings are the results concerning DCH loads in descending order. 1. Availability of ignition sources prior to vessel breach 2. availability and effectiveness of ice in the ice condenser 3. Loads modeling uncertainties related to co-ejected RPV water 4. Other loads modeling uncertainties 10 tabs., 3 figs., 14 refs. (Author)

  18. DCH analyses using the CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Wan; Kim, Hee Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-08-01

    This report describes CONTAIN analyses performed during participation in the project of `DCH issue resolution for ice condenser plants` which is sponsored by NRC at SNL. Even though the calculations were performed for the Ice Condenser plant, CONTAIN code has been used for analyses of many phenomena in the PWR containment and the DCH module can be commonly applied to any plant types. The present ice condenser issue resolution effort intended to provide guidance as to what might be needed to resolve DCH for ice condenser plants. It includes both a screening analysis and a scoping study if the screening analysis cannot provide an complete resolution. The followings are the results concerning DCH loads in descending order. 1. Availability of ignition sources prior to vessel breach 2. availability and effectiveness of ice in the ice condenser 3. Loads modeling uncertainties related to co-ejected RPV water 4. Other loads modeling uncertainties 10 tabs., 3 figs., 14 refs. (Author).

  19. IDIOMS CONTAINING THE COMPONENT BLACK / SCHWARZ IN THE ENGLISH AND GERMAN LANGUAGES: COMPARATIVE ANALYSIS

    Directory of Open Access Journals (Sweden)

    Yakovleva, S.L.

    2016-06-01

    Full Text Available The article presents a comparative analysis of English and German idiomatic expressions containing the component of colour in their structure. It has been revealed that black dominates in the English linguistic idiomatic view of the world. The core centre of the focal colours in the German culture is also schwarz / black. General and specific features of black / schwarz as a part of national linguistic views of the world of the English and German languages are considered in the article.

  20. NAUAHYGROS - A code for calculating aerosol behavior in nuclear power plant containments following a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)

    1995-02-01

    NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.

  1. 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

  2. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  3. A computer code to design liquid containers for vehicles

    International Nuclear Information System (INIS)

    Parizi, H.B.; Fard, M.P.; Dolatabadi, A.

    2003-01-01

    We are presenting the development of a modular code for the simulation of the fluid sloshing that occurs in the liquid containers in vehicles. Sloshing occurs when a partially filled container of liquid goes through transient or steady external forces. Under such conditions, the free surface of the liquid may move and the liquid may impact on the walls of the container, exchanging forces. These forces may cause numerous harmful and undesirable consequences in the operation of the vehicle, such as vehicle turn over. The fluid mechanic equations that describe the fluid sloshing in the container and the dynamic equations that describe the movement of the container are solved separately in two different codes. The codes are coupled weekly, such that the output of one code will be used as the input to the other code in the same time step. The outputs of the fluid code are the forces and torques that are applied to the body of the container due to sloshing, whereas the output of the dynamic code are the translational and rotational velocities and accelerations of the container. The proposed software can be used to test the performance of the designed container under various operating condition and allow effective improvements to the container design. The proposed code is different than the presently available codes, in that it will provide a true simulation of the coupled fluid and structure interaction. (author)

  4. 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

  5. Recent developments in the CONTAIN-LMR code

    International Nuclear Information System (INIS)

    Murata, K.K.

    1990-01-01

    Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab

  6. A development of containment performance analysis methodology using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. C.; Yoon, J. I. [Future and Challenge Company, Seoul (Korea, Republic of); Byun, C. S.; Lee, J. Y. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, J. Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-10-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code.

  7. A development of containment performance analysis methodology using GOTHIC code

    International Nuclear Information System (INIS)

    Lee, B. C.; Yoon, J. I.; Byun, C. S.; Lee, J. Y.; Lee, J. Y.

    2003-01-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code

  8. The WINCON programme - validation of fast reactor primary containment codes

    International Nuclear Information System (INIS)

    Sidoli, J.E.A.; Kendall, K.C.

    1988-01-01

    In the United Kingdom safety studies for the Commercial Demonstration Fast Reactor (CDFR) include an assessment of the capability of the primary containment in providing an adequate containment for defence against the hazards resulting from a hypothetical Whole Core Accident (WCA). The assessment is based on calculational estimates using computer codes supported by measured evidence from small-scale experiments. The hydrodynamic containment code SEURBNUK-EURDYN is capable of representing a prescribed energy release, the sodium coolant and cover gas, and the main containment and safety related internal structures. Containment loadings estimated using SEURBNUK-EURDYN are used in the structural dynamic code EURDYN-03 for the prediction of the containment response. The experiments serve two purposes, they demonstrate the response of the CDFR containment to accident loadings and provide data for the validation of the codes. This paper summarises the recently completed WINfrith CONtainment (WINCON) experiments that studied the response of specific features of current CDFR design options to WCA loadings. The codes have been applied to some of the experiments and a satisfactory prediction of the global response of the model containment is obtained. This provides confidence in the use of the codes in reactor assessments. (author)

  9. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper describes the application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third problem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed. (Auth.)

  10. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    The application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs is described. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third proem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed

  11. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together

  12. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  13. Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kwi Hyun; Song, Wan Jung [ENERGEO Inc. Sungnam, (Korea, Republic of); Song, Dong Soo; Byun, Choong Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences.

  14. Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

    International Nuclear Information System (INIS)

    Seo, Kwi Hyun; Song, Wan Jung; Song, Dong Soo; Byun, Choong Sup

    2006-01-01

    The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences

  15. Liquid metal reactor applications of the CONTAIN code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Gido, R.; Valdez, G.D.; Scholtyssek, W.

    1988-01-01

    The CONTAIN code is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during a severe accident. Included in the phenomena modeled are thermal-hydraulics, radiant and convective heat transfer, aerosol loading and transient response, fission product transport and heating effects, and interactions of sodium and corium with the containment atmosphere and structures. CONTAIN has been used by groups in Japan and West Germany to assess its ability to analyze accident consequences for liquid metal reactor (LMR) plants. In conjunction with this use, collaborative efforts to improve the modeling have been pursued. This paper summarizes the current state of the version of CONTAIN that has been enhanced with extra capabilities for LMR applications. A description of physical models is presented, followed by a review of validation exercises performed with CONTAIN. Some demonstration calculations of an integrated LMR application are presented

  16. Is Self-Regulation Sufficient? Case of the German Transparency Code

    Directory of Open Access Journals (Sweden)

    Kristin Buske

    2016-02-01

    Full Text Available The German pharmaceutical industry is stepping ahead with its implementation of a new transparency disclosure code for cooperation between pharmaceutical companies and health care professionals (HCPs and health care organisations (HCOs. In Germany, this transparency code (“Transparenzkodex” is applicable since January 2015, and data will be publicly available around mid-2016. No empirical work has been done that addresses the impact of the transparency code on cooperation between HCPs, HCOs and the pharmaceutical companies, including the possibilities of competitive analysis of the available data. In this paper, we interviewed experts from 11 pharmaceutical companies representing small, medium-sized as well as multinational corporations which represent 80% of the German pharmaceutical market. Besides interviews, the authors designed a game to evaluate possible financial investments in key opinion leaders. The market can be regarded as a zero sum game. By allowing public identification of such key HCPs and HCOs, the amount spent on them might increase and not decrease. In a way, the transparency code may foster more and not less spending; in our simulation game, the financial investment in marketing key HCPs and HCOs exceeded sustainable limits.

  17. German nuclear codes revised: comparison with approaches used in other countries

    International Nuclear Information System (INIS)

    Raetzke, C.; Micklinghoff, M.

    2005-01-01

    The article deals with the plan of the German Federal Ministry for the Environment (BMU) to revise the German set of nuclear codes, and draws a comparison with approaches pursued in other countries in formulating and implementing new requirements imposed upon existing plants. A striking feature of the BMU project is the intention to have the codes reflect the state of the art in an entirely abstract way irrespective of existing plants. This implies new requirements imposed on plant design, among other things. However, the state authorities, which establish the licensing conditions for individual plants in concrete terms, will not be able to apply these new codes for legal reasons (protection of vested rights) to the extent in which they incorporate changes in safety philosophy. Also the procedure adopted has raised considerable concern. The processing time of two years is inordinately short, and participation of the public and of industry does not go beyond the strictly formal framework of general public participation. In the light of this absence of quality assurance, it would be surprising if this new set of codes did not suffer from considerable deficits in its contents. Other countries show that the BMU is embarking on an isolated approach in every respect. Elsewhere, backfitting requirements are developed carefully and over long periods of time; they are discussed in detail with the operators; costs and benefits are weighted, and the consequences are evaluated. These elements are in common to procedures in all countries, irrespective of very different steps in detail. (orig.)

  18. LWR containment thermal hydraulic codes benchmark demona B3 exercise

    International Nuclear Information System (INIS)

    Della Loggia, E.; Gauvain, J.

    1988-01-01

    Recent discussion about the aerosol codes currently used for the analysis of containment retention capabilities have revealed a number of questions concerning the reliabilities and verifications of the thermal-hydraulic modules of these codes with respect to the validity of implemented physical models and the stability and effectiveness of numerical schemes. Since these codes are used for the calculation of the Source Term for the assessment of radiological consequences of severe accidents, they are an important part of reactor safety evaluation. For this reason the Commission of European Communities (CEC), following the recommendation mode by experts from Member Stades, is promoting research in this field with the aim also of establishing and increasing collaboration among Research Organisations of member countries. In view of the results of the studies, the CEC has decided to carry out a Benchmark exercise for severe accident containment thermal hydraulics codes. This exercise is based on experiment B3 in the DEMONA programme. The main objective of the benchmark exercise has been to assess the ability of the participating codes to predict atmosphere saturation levels and bulk condensation rates under conditions similar to those predicted to follow a severe accident in a PWR. This exercise follows logically on from the LA-4 exercise, which, is related to an experiment with a simpler internal geometry. We present here the results obtained so far and from them preliminary conclusions are drawn, concerning condensation temperature, pressure, flow rates, in the reactor containment

  19. Validation of containment thermal hydraulic computer codes for VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

    2005-07-01

    Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

  20. The Mistra experiment for field containment code validation first results

    International Nuclear Information System (INIS)

    Caron-Charles, M.; Blumenfeld, L.

    2001-01-01

    The MISTRA facility is a large scale experiment, designed for the purpose of thermal-hydraulics multi-D codes validation. A short description of the facility, the set up of the instrumentation and the test program are presented. Then, the first experimental results, studying helium injection in the containment and their calculations are detailed. (author)

  1. Experimental validation of the containment codes ASTARTE and SEURBNUK

    International Nuclear Information System (INIS)

    Kendall, K.C.; Arnold, L.A.; Broadhouse, B.J.; Jones, A.; Yerkess, A.; Benuzzi, A.

    1979-10-01

    The fast reactor containment codes ASTARTE and SEURBNUK are being validated against data from the COVA series of small scale experiments being performed jointly by the UKAEA and JRC Ispra. The experimental programme is nearly complete, and data are given. (U.K.)

  2. Comparison of ANL containment codes with SNR-300 simulation experiments

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Wang, C.Y.; Fistedis, S.H.

    1976-01-01

    A comparison of REXCO and ICECO code predictions is made with data obtained from experiments of LMFBR excursion models. The comparisons are based on published results of tests conducted for the safety analysis of the SNR-300 fast breeder. The test configurations consist of a centrally located spherical source immersed in a pool of water which is encased in a cylindrical container. The cylinical walls of the container are prestressed by holddown bolts which span the two rigid ends. The space above the surface of the water within the container is occupied by air. Although certain aspects of the tests could not be simulated by the analytical models exactly, the comparison of results shows quite close agreement. The fact that the REXCO and ICECO codes involve different analytical formulations, their own close correspondence of results lends added credence to the value of analytical predictions

  3. GOTHIC code evaluation of alternative passive containment cooling features

    International Nuclear Information System (INIS)

    Gavrilas, M.; Todreas, E.N.; Driscoll, M.J.

    1996-01-01

    Reliance on passive cooling has become an important objective in containment design. Several reactor concepts have been set forth, which are equipped with entirely passively cooled containments. However, the problems that have to be overcome in rejecting the entire heat generated by a severe accident in a high-rating reactor (i.e. one with a rating greater than 1200 MW e ) have been found to be substantial and without obvious solutions. The GOTHIC code was verified and modified for containment cooling applications; optimal mesh sizes, computational time steps and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. The GOTHIC code was then employed to assess the effectiveness of several original heat rejection features that make it possible to cool high-rating containments. Two containment concepts were evaluated: one for a 1200 MW e new pressure tube light-water reactor, and one for a 1300 MW e pressurized-water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features has been predicted. The best-performance configurations-worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light-water reactor, and less than 0.45 MPa for the 1300 MW e pressurized-water reactor. (orig.)

  4. Die Kodifikation des deutschen Nichtehelichenrechts im Bürgerlichen Gesetzbuch The Codification of German Non-Marriage Law in the German Civil Code

    Directory of Open Access Journals (Sweden)

    Eric Neiseke

    2008-07-01

    Full Text Available Steffen Baumgarten legt erstmals eine umfassende Darstellung zur Kodifikation des Nichtehelichenrechts im Bürgerlichen Gesetzbuch unter Berücksichtigung der Stellungnahmen der deutschen Frauenbewegung vor. Zugleich werden die sozialen und gesellschaftlichen Hintergründe im 19. Jahrhundert in die Untersuchung mit einbezogen.Steffen Baumgarten presents the first comprehensive presentation of the codification of “non-marriage laws” in the German Civil Code in light of the position of the German women’s movement. His study also includes the social and societal background of the 19th century.

  5. Ice condenser containment analysis with the GOTHIC code

    International Nuclear Information System (INIS)

    Yadon, T.P.

    1996-01-01

    Analytical methodologies have recently been developed by Duke Power Company (Duke) to calculate the thermodynamic response of the ice condenser containment buildings at the McGuire and Catawba Nuclear Stations to high-energy line breaks. The GOTHIC computer code (Version 4.0) was utilized for these analyses. In the ice condenser containment design, a large mass of ice stored within the containment building is used to absorb the energy released from high-energy line breaks, thereby limiting the peak pressure and temperature in the containment building to within design limits. The McGuire and Catawba Nuclear Stations (both two-unit, 3411 MWth four-loop Westinghouse plants) are of the ice condenser containment design

  6. Container code recognition in information auto collection system of container inspection

    International Nuclear Information System (INIS)

    Su Jianping; Chen Zhiqiang; Zhang Li; Gao Wenhuan; Kang Kejun

    2003-01-01

    Now custom needs electrical application and automatic detection. Container inspection should not only give the image of the goods, but also auto-attain container's code and weight. Its function and track control, information transfer make up the Information Auto Collection system of Container Inspection. Code Recognition is the point. The article is based on model match, the close property of character, and uses it to recognize. Base on checkout rule, design the adjustment arithmetic, form the whole recognition strategy. This strategy can achieve high recognition ratio and robust property

  7. GOTHIC code evaluation of alternative passive containment cooling features

    International Nuclear Information System (INIS)

    Gavrilas, M.; Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

    1994-01-01

    The GOTHIC code was employed to assess the effectiveness of several original heat rejection features that make it possible to cool large rating containments. The code was first verified and modified for specific containment cooling applications; optimal mesh sizes, computational time steps, and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. GOTHIC was then used to obtain performance predictions for two containment concepts: a 1200 MW e new pressure tube light water reactor, and a 1300 MW e pressurized water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features have been predicted. For the 1200 MW e pressure tube light water reactor, the evaluated pressure-limiting features are: a large water pool connected to the calandria, large containment free volume and an air-convection annulus. For the 1300 MW e pressurized water reactor, an external moat, an internal water pool, and an air-convection annulus were evaluated. The performance of the proposed containment configurations is dependent on the extent of thermal stratification inside the containment. The best-performance configurations/worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light water reactor, and less than 0.45 MPa for the 1300 MW e pressurized water reactor. The low peak pressure predicted for the 1200 MW e pressure tube light water reactor can be in part attributed to its relatively large free volume, while the relatively high peak pressure predicted for the 1300 MW e pressurized water reactor can be attributed to its relatively small free volume (i.e., the size used was that of a pressurized water reactor containment designed with active heat removal features). (author)

  8. Probabilistic analysis of crack containing structures with the PARIS code

    International Nuclear Information System (INIS)

    Brueckner-Foit, A.

    1987-10-01

    The basic features of the PARIS code which has been developed for the calculation of failure probabilities of crack containing structures are explained. An important issue in the reliability analysis of cracked components is the probabilistic leak-before-break behaviour. Formulae for the leak and break probabilities are derived and it is shown how a leak detection system influences the results. An example taken from nuclear applications illustrates the details of the probabilistic leak-before-break analysis. (orig.) [de

  9. PCCS model development for SBWR using the CONTAIN code

    International Nuclear Information System (INIS)

    Tills, J.; Murata, K.K.; Washington, K.E.

    1994-01-01

    The General Electric Simplified Boiling Water Reactor (SBWR) employs a passive containment cooling system (PCCS) to maintain long-term containment gas pressure and temperature below design limits during accidents. This system consists of a steam supply line that connects the upper portion of the drywell with a vertical shell-and-tube single pass heat exchanger located in an open water pool outside of the containment safety envelope. The heat exchanger tube outlet is connected to a vent line that is submerged below the suppression pool surface but above the main suppression pool horizontal vents. Steam generated in the post-shutdown period flows into the heat exchanger tubes as the result of suction and/or a low pressure differential between the drywell and suppression chamber. Operation of the PCCS is complicated by the presence of noncondensables in the flow stream. Build-up of noncondensables in the exchanger and vent line for the periods when the vent is not cleared causes a reduction in the exchanger heat removal capacity. As flow to the exchanger is reduced due to the noncondensable gas build-up, the drywell pressure increases until the vent line is cleared and the noncondensables are purged into the suppression chamber, restoring the heat removal capability of the PCCS. This paper reports on progress made in modeling SBWR containment loads using the CONTAIN code. As a central part of this effort, a PCCS model development effort has recently been undertaken to implement an appropriate model in CONTAIN. The CONTAIN PCCS modeling approach is discussed and validated. A full SBWR containment input deck has also been developed for CONTAIN. The plant response to a postulated design basis accident (DBA) has been calculated with the CONTAIN PCCS model and plant deck, and the preliminary results are discussed

  10. Modeling of hydrogen behaviour in a PWR nuclear power plant containment with the CONTAIN code

    International Nuclear Information System (INIS)

    Bobovnik, G.; Kljenak, I.

    2001-01-01

    Hydrogen behavior in the containment during a severe accident in a two-loop Westinghouse-type PWR nuclear power plant was simulated with the CONTAIN code. The accident was initiated with a cold-leg break of the reactor coolant system in a steam generator compartment. In the input model, the containment is represented with 34 cells. Beside hydrogen concentration, the containment atmosphere temperature and pressure and the carbon monoxide concentration were observed as well. Simulations were carried out for two different scenarios: with and without successful actuation of the containment spray system. The highest hydrogen concentration occurs in the containment dome and near the hydrogen release location in the early stages of the accident. Containment sprays do not have a significant effect on hydrogen stratification.(author)

  11. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  12. How L2-Learners' Brains React to Code-Switches: An ERP Study with Russian Learners of German

    Science.gov (United States)

    Ruigendijk, Esther; Hentschel, Gerd; Zeller, Jan Patrick

    2016-01-01

    This Event Related Potentials (ERP) study investigates auditory processing of sentences with so-called code-switches in Russian learners of German. It has often been argued that switching between two languages results in extra processing cost, although it is not completely clear yet what exactly causes these costs. ERP presents a good method to…

  13. Legal and economic implications of the German Craft Code Amendment 2004

    Directory of Open Access Journals (Sweden)

    Wolfgang Benzel

    2012-01-01

    Full Text Available As the main reason for initiating an amendment of the Trade and Crafts Code in 2004 was the decrease in the number of companies in general and the number of skilled crafts enterprises in particular and was also due to the fact that the number of employees in the skilled crafts sector had constantly fallen and many companies were not able to find a successor, it was the declared aim of the amendment to counter this structural crisis as well as reducing illegal employment and removing native discrimination. This would then ensure a higher level of employment especially through new company formations and takeovers of existing companies.In addition to providing an important contribution towards the assessment of the basic success of the 2004 Trade and Crafts Code amendment, the results can therefore be used at an international level for discussions on skilled crafts legislation outside Germany. As German skilled crafts legislation developed in a completely different manner from skilled crafts legislation in most other European countries, this aspect is particularly significant and it is conceivable that the results will be useful beyond the borders of the Federal Republic of Germany. The research findings will also be important in determining the limits of deregulation and when it can be considered counterproductive from an economic point of view.

  14. Simulation of containment phenomena during the Phebus FPT1 test with the CONTAIN code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2002-01-01

    Thermal-hydraulic and aerosol phenomena which occurred in the containment vessel of the Phebus integral experimental facility during the first 30000 s of the Phebus FPT1 test were simulated with the CONTAIN thermal-hydraulic computer code. A single-cell input model of the vessel was developed, and boundary and initial conditions that were determined during the experiment were applied. The comparison of experimental and calculated results shows that, although the atmosphere temperature was well simulated, the calculated condensation rate was apparently too high, resulting in a lower pressure of the containment atmosphere. The aerosol deposition process was well simulated.(author)

  15. Modeling of hydrogen stratification in a pressurized water reactor containment with the contain computer code

    International Nuclear Information System (INIS)

    Kljenak, I.; Skerlavaj, A.; Parzer, I.

    1999-01-01

    Hydrogen distribution during a severe accident in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN computer code. The accidents is initiated by a large-break loss-of-coolant accident which is nit successfully mitigated by the action of the emergency core cooling system. Cases with and without successful actuation of spray systems and fan coolers were considered. The simulations predicted hydrogen stratification within the containment main compartment with intensive hydrogen mixing in the containment dome region. Pressure and temperature responses were analyzed as well.(author)

  16. Hydrogen distribution analysis for CANDU 6 containment using the GOTHIC containment analysis code

    International Nuclear Information System (INIS)

    Nguyen, T.H.; Collins, W.M.

    1995-01-01

    Hydrogen may be generated in the reactor core by the zircaloy-steam reaction for a postulated loss of coolant accident (LOCA) scenario with loss of emergency core cooling (ECC). It is important to predict hydrogen distribution within containment in order to determine if flammable mixtures exist. This information is required to determine the best locations in containment for the placement of mitigation devices such as igniters and recombiners. For large break loss coolant accidents, hydrogen is released after the break flow has subsided. Following this period of high discharge the flow in the containment building undergoes transition from forced flow to a buoyancy driven flow (particularly when local air coolers (LACS) are not credited). One-dimensional computer codes (lumped parameter) are applicable during the initial period when a high degree of mixing occurs due to the forced flow generated by the break. However, during the post-blowdown phase the assumption of homogeneity becomes less accurate, and it is necessary to employ three-dimensional codes to capture local effects. This is particularly important for purely buoyant flows which may exhibit stratification effects. In the present analysis a three-dimensional model of CANDU 6 containment was constructed with the GOTHIC computer code using a relatively coarse mesh adequate enough to capture the salient features of the flow during the blowdown and hydrogen release periods. A 3D grid representation was employed for that portion of containment in which the primary flow (LOCA and post-LOCA) was deemed to occur. The remainder of containment was represented by lumped nodes. The results of the analysis indicate that flammable concentrations exist for several minutes in the vicinity of the break and in the steam generator enclosure. This is due to the fact that the hydrogen released from the break is primarily directed upwards into the steam generator enclosure due to buoyancy effects. Once hydrogen production ends

  17. Effective modeling of hydrogen mixing and catalytic recombination in containment atmosphere with an Eulerian Containment Code

    International Nuclear Information System (INIS)

    Bott, E.; Frepoli, C.; Monti, R.; Notini, V.; Carcassi, M.; Fineschi, F.; Heitsch, M.

    1999-01-01

    Large amounts of hydrogen can be generated in the containment of a nuclear power plant following a postulated accident with significant fuel damage. Different strategies have been proposed and implemented to prevent violent hydrogen combustion. An attractive one aims to eliminate hydrogen without burning processes; it is based on the use of catalytic hydrogen recombiners. This paper describes a simulation methodology which is being developed by Ansaldo, to support the application of the above strategy, in the frame of two projects sponsored by the Commission of the European Communities within the IV Framework Program on Reactor Safety. Involved organizations also include the DCMN of Pisa University (Italy), Battelle Institute and GRS (Germany), Politechnical University of Madrid (Spain). The aims to make available a simulation approach, suitable for use for containment design at industrial level (i.e. with reasonable computer running time) and capable to correctly capture the relevant phenomenologies (e.g. multiflow convective flow patterns, hydrogen, air and steam distribution in the containment atmosphere as determined by containment structures and geometries as well as by heat and mass sources and sinks). Eulerian algorithms provide the capability of three dimensional modelling with a fairly accurate prediction, however lower than CFD codes with a full Navier Stokes formulation. Open linking of an Eulerian code as GOTHIC to a full Navier Stokes CFD code as CFX 4.1 allows to dynamically tune the solving strategies of the Eulerian code itself. The effort in progress is an application of this innovative methodology to detailed hydrogen recombination simulation and a validation of the approach itself by reproducing experimental data. (author)

  18. Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code

    International Nuclear Information System (INIS)

    Williams, D.C.; Bergeron, K.D.; Carroll, D.E.; Gasser, R.D.; Tills, J.L.; Washington, K.E.

    1987-05-01

    One of the most important unresolved issues governing risk in many nuclear power plants involves the phenomenon called direct containment heating (DCH), in which it is postulated that molten corium ejected under high pressure from the reactor vessel is dispersed into the containment atmosphere, thereby causing sufficient heating and pressurization to threaten containment integrity. Models for the calculation of potential DCH loads have been developed and incorporated into the CONTAIN code for severe accident analysis. Using CONTAIN, DCH scenarios in PWR plants having three different representative containment types have been analyzed: Surry (subatmospheric large dry containment), Sequoyah (ice condenser containment), and Bellefonte (atmospheric large dry containment). A large number of parameter variation and phenomenological uncertainty studies were performed. Response of DCH loads to these variations was found to be quite complex; often the results differ substantially from what has been previously assumed concerning DCH. Containment compartmentalization offers the potential of greatly mitigating DCH loads relative to what might be calculated using single-cell representations of containments, but the actual degree of mitigation to be expected is sensitive to many uncertainties. Dominant uncertainties include hydrogen combustion phenomena in the extreme environments produced by DCH scenarios, and factors which affect the rate of transport of DCH energy to the upper containment. In addition, DCH loads can be aggravated by rapid blowdown of the primary system, co-dispersal of moderate quantities of water with the debris, and quenching of de-entrained debris in water; these factors act by increasing steam flows which, in turn, accelerates energy transport. It may be noted that containment-threatening loads were calculated for a substantial portion of the scenarios treated for some of the plants considered

  19. Comparative study of design of piping supports class 1, 2 and 3 considering german code KTA and ASME III - NF

    International Nuclear Information System (INIS)

    Faloppa, Altair A.; Fainer, Gerson; Mattar Neto, Miguel; Elias, Marcos V.

    2013-01-01

    The objective of this paper is developing a comparative study of the design criteria for class 1, 2, 3 piping supports considering the American Code ASME Section III - NF and the German Code KTA 3205.1 to the Primary Circuit, KTA 3205.2 to the others systems and KTA 3205.3 series-production standards supports of a PWR nuclear power plant. An additional purpose of the paper is a general analysis of the main design concepts of the American Code ASME Boiler and Pressure Vessel Code, Section III, Division 1 and German Nuclear Design Code KTA that was performed in order to aid the comparative study proposed. The relevance of this study is to show the differences between codes ASME and KTA since they were applied in the design of the Nuclear Power Plants Angra 1 and Angra 2, and to the design of Angra 3, which is at the moment under construction. It is also considered their use in the design of nuclear installations such as RMB - Reator MultiProposito Brasileiro and LABGENE - Laboratorio de Geracao Nucleoeletrica. (author)

  20. [Quality management and strategic consequences of assessing documentation and coding under the German Diagnostic Related Groups system].

    Science.gov (United States)

    Schnabel, M; Mann, D; Efe, T; Schrappe, M; V Garrel, T; Gotzen, L; Schaeg, M

    2004-10-01

    The introduction of the German Diagnostic Related Groups (D-DRG) system requires redesigning administrative patient management strategies. Wrong coding leads to inaccurate grouping and endangers the reimbursement of treatment costs. This situation emphasizes the roles of documentation and coding as factors of economical success. The aims of this study were to assess the quantity and quality of initial documentation and coding (ICD-10 and OPS-301) and find operative strategies to improve efficiency and strategic means to ensure optimal documentation and coding quality. In a prospective study, documentation and coding quality were evaluated in a standardized way by weekly assessment. Clinical data from 1385 inpatients were processed for initial correctness and quality of documentation and coding. Principal diagnoses were found to be accurate in 82.7% of cases, inexact in 7.1%, and wrong in 10.1%. Effects on financial returns occurred in 16%. Based on these findings, an optimized, interdisciplinary, and multiprofessional workflow on medical documentation, coding, and data control was developed. Workflow incorporating regular assessment of documentation and coding quality is required by the DRG system to ensure efficient accounting of hospital services. Interdisciplinary and multiprofessional cooperation is recognized to be an important factor in establishing an efficient workflow in medical documentation and coding.

  1. 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

  2. Aspects of a generic photovoltaic model examined under the German grid code for medium voltage

    Energy Technology Data Exchange (ETDEWEB)

    Theologitis, Ioannis-Thomas; Troester, Eckehard; Ackermann, Thomas [Energynautics GmbH, Langen (Germany)

    2011-07-01

    The increasing peneration of photovoltaic power systems into the power grid has attached attention to the issue of ensuring the smooth absorbance of the solar energy, while securing the normal and steady operation of the grid as well. Nowadays, the PV systems must meet a number of technical requirements to address this issue. This paper investigates a generic grid-connected photovoltaic model that was developed by DIgSILENT and is part of the library in the new version of PowerFactory v.14.1 software that is used in this study. The model has a nominal rated peak power of 0.5 MVA and a designed power factor cos{phi}0.95. The study focuses on the description of the model, its control system and its ability to reflect important requirements that a grid-connected PV system should have by January 2011 according to the German grid code for medium voltage. The model undergoes various simulations. Static voltage support, active power control and dynamic voltage support - Fault Ride Through (FRT) is examined. The results show that the generic model is capable for active power reduction under over-frequency occasions and FRT behavior in cases of voltage dips. The reactive power control that is added in the model improves the control system and makes the model capable for static voltage support in sudden active power injection changes at the point of common coupling. Beside the simplifications and shortcomings of this generic model, basic requirements of the modern PV systems can be addressed. Further improvements could make it more complete and applicable for more detailed studies. (orig.)

  3. ZERBERUS - the code for reliability analysis of crack containing structures

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.

    1992-04-01

    Brief description of the First- and Second Order Reliability Methods, being the theoretical background of the code, is given. The code structure is described in detail, with special emphasis to the new application fields. The numerical example investigates failure probability of steam generator tubing affected by stress corrosion cracking. The changes necessary to accommodate this analysis within the ZERBERUS code are explained. Analysis results are compared to different Monte Carlo techniques. (orig./HP) [de

  4. Development and validation of a catalytic recombiner model for the containment code RALOC MOD4.0

    International Nuclear Information System (INIS)

    Rohde, J.; Klein-Hebling, W.; Chakraborty, A.K.

    1997-01-01

    This paper reports on the development of a catalytic recombiner model for the containment code RALOC MOD4.0 /KLH 95, KLH 96/ and the detailed validation work, carried out at GRS. The model was qualified by using the results of medium and large scale experiments, being performed in Germany /KAN 91/. The comparison of measured data with the calculations demonstrates, that this new model is suitable for real plant applications to investigate the overall effectiveness of a catalytic recombiner system under severe accident conditions for large dry containments of German PWR design. The results of such investigations will serve as the basis to work out some guidance for the determination of the system capacity needed and an optimal positioning of such devices in containments. (author)

  5. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  6. Code of Practice containing definitions for Safety Codes of Practice for nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    This Code provides definitions of the technical terms used in the licensing applications to be submitted to the Turkish Atomic Energy Commission (TAEC), in accordance with national licensing regulations. The Code is based mainly on the International Atomic Energy Agency's Code of Practice on the subject. (NEA) [fr

  7. Application of the MELCOR code to design basis PWR large dry containment analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  8. CONTAIN 2.0 code release and the transition to licensing

    International Nuclear Information System (INIS)

    Murata, K.K.; Griffith, R.O.; Bergeron, K.D.; Tills, J.

    1997-10-01

    CONTAIN is a reactor accident simulation code developed by Sandia National Laboratories under US Nuclear Regulatory Commission (USNRC) sponsorship to provide integrated analysis of containment phenomena, including those related to nuclear reactor containment loads and radiological source terms. The recently released CONTAIN 2.0 code version represents a significant advance in CONTAIN modeling capabilities over the last major code release (CONTAIN 1.12V). The new modeling capabilities are discussed here. The principal motivation for many of the recent model improvements has been to allow CONTAIN to model the special features in advanced light water reactor (ALWR) designs. The work done in this area is also summarized. In addition to the ALWR work, the USNRC is currently engaged in an effort to qualify CONTAIN for more general use in licensing, with the intent of supplementing or possibly replacing traditional licensing codes. To qualify the CONTAIN code for licensing applications, studies utilizing CONTAIN 2.0 are in progress. A number of results from this effort are presented in this paper to illustrate the code capabilities. In particular, CONTAIN calculations of the NUPEC M-8-1 and ISP-23 experiments and CVTR test number-sign 3 are presented to illustrate (1) the ability of CONTAIN to model non-uniform gas density and/or temperature distributions, and (2) the relationship between such gas distributions and containment loads. CONTAIN and CONTEMPT predictions for a large break loss of coolant accident scenario in the San Onofre plant are also compared

  9. Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

    International Nuclear Information System (INIS)

    Risken, Tobias; Koch, Marco K.

    2011-01-01

    The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

  10. Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1986-01-01

    The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models

  11. Are industry codes and standards a valid cost containment approach

    International Nuclear Information System (INIS)

    Rowley, C.W.; Simpson, G.T.; Young, R.K.

    1990-01-01

    The nuclear industry has historically concentrated on safety design features for many years, but recently has been shifting to the reliability of the operating systems and components. The Navy has already gone through this transition and has found that Reliability Centered Maintenance (RCM) is an invaluable tool to improve the reliability of components, systems, ships, and classes of ships. There is a close correlation of Navy ships and equipment to commercial nuclear power plants and equipment. The Navy has a central engineering and configuration management organization (Naval Sea Systems Command) for over 500 ships, where as the over 100 commercial nuclear power plants and 52 nuclear utilities represent a fragmented owner/management structure. This paper suggests that the results of the application of RCM in the Navy can be duplicated to a large degree in the commercial nuclear power industry by the development and utilization of nuclear codes and standards

  12. Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

    International Nuclear Information System (INIS)

    Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

    2010-01-01

    Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

  13. A research on verification of the CONTAIN CODE model and the uncertainty reduction method for containment integrity

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae-Hong; Kim, Moo-Hwan; Bae, Seong-Won; Byun, Sang-Chul [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    1998-03-15

    The final objectives of this study are to establish the way of measuring the integrity of containment building structures and safety analysis in the period of a postuIated severe accidents and to decrease the uncertainty of these methods. For that object, the CONTAIN 1.2 codes model for analyzing the severe accidents phenomena and the heat transfer between the air inside the containment buildings and inner walls have been reviewed and analyzed. For the double containment wall provided to the next generation nuclear reactor, which is different to the previous type of containment, the temperature and pressure rising history were calculated and compared to the results of previous ones.

  14. The assessment of containment codes by experiments simulating severe accident scenarios

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-01-01

    Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

  15. Reliability-based design code calibration for concrete containment structures

    International Nuclear Information System (INIS)

    Han, B.K.; Cho, H.N.; Chang, S.P.

    1991-01-01

    In this study, a load combination criteria for design and a probability-based reliability analysis were proposed on the basis of a FEM-based random vibration analysis. The limit state model defined for the study is a serviceability limit state of the crack failure that causes the emission of radioactive materials, and the results are compared with the case of strength limit state. More accurate reliability analyses under various dynamic loads such as earthquake loads were made possible by incorporating the FEM and random vibration theory, which is different from the conventional reliability analysis method. The uncertainties in loads and resistance available in Korea and the references were adapted to the situation of Korea, and especially in case of earthquake, the design earthquake was assessed based on the available data for the probabilistic description of earthquake ground acceleration in the Korea peninsula. The SAP V-2 is used for a three-dimensional finite element analysis of concrete containment structure, and the reliability analysis is carried out by modifying HRAS reliability analysis program for this study. (orig./GL)

  16. Development of CAP code for nuclear power plant containment: Lumped model

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

    2015-09-15

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.

  17. Impact of ACI-ASME code on design and construction of nuclear containment structures

    International Nuclear Information System (INIS)

    Reedy, R.F.

    1978-01-01

    The effect of the ACI-ASME code for design and construction of concrete containment structures on the nuclear and concrete industries is examined. Topics covered include purpose of the code, general requirements, responsibilities and duties, design and construction specifications, quality assurance, inspection, the liner, and stamping

  18. Development of CAP code for nuclear power plant containment: Lumped model

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun

    2015-01-01

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP

  19. Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.; Gvildys, J.

    1977-01-01

    In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs

  20. Calculation of behaviour of the Juragua NPP containment with code TRACOV/MOD1

    International Nuclear Information System (INIS)

    Castillo Alvarez, J.; Valle Cepero, R.; Luis, J.; San Roman, J.C.; Pomier, L.

    1996-01-01

    The containment of Juragua NPP has some unique features, which differ from the rest of the PWR reactors design. Those features impose additional requirements for its numerical simulation. In this paper is analyzed the behaviour of the Juragua NPP containment during accident situation with double ended break of the primary pipelines with flow in both direction using the code TRACOV/MOD1. The results are compared with obtained by the designer. The main restrictions of the code are identified

  1. Integrated analysis of core debris interactions and their effects on containment integrity using the CONTAIN computer code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Williams, D.C.; Tills, J.L.; Valdez, G.D.

    1987-01-01

    The CONTAIN computer code includes a versatile system of phenomenological models for analyzing the physical, chemical and radiological conditions inside the containment building during severe reactor accidents. Important contributors to these conditions are the interactions which may occur between released corium and cavity concrete. The phenomena associated with interactions between ejected corium debris and the containment atmosphere (Direct Containment Heating or DCH) also pose a potential threat to containment integrity. In this paper, we describe recent enhancements of the CONTAIN code which allow an integrated analysis of these effects in the presence of other mitigating or aggravating physical processes. In particular, the recent inclusion of the CORCON and VANESA models is described and a calculation example presented. With this capability CONTAIN can model core-concrete interactions occurring simultaneously in multiple compartments and can couple the aerosols thereby generated to the mechanistic description of all atmospheric aerosol components. Also discussed are some recent results of modeling the phenomena involved in Direct Containment Heating. (orig.)

  2. Code for calculation of spreading of radioactivity in reactor containment systems

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

    A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs

  3. Best estimate procedures for fatigue evaluation in the framework of German KTA code

    Energy Technology Data Exchange (ETDEWEB)

    Seichter, Johannes [Siempelkamp Pruef- und Gutachter-Gesellschaft mbH, Dresden (Germany); Reese, Sven H.; Klucke, Dietmar [E.ON Kernkraft GmbH, Hannover (Germany)

    2013-07-01

    By decreasing the level of conservatism in fatigue analyses it is possible to reduce as well fatigue usage factors calculated for EOL (end of life) as 'actual CUF' (cumulative fatigue usage factor) of NPP components considerably. It is the opinion of the authors, that the mentioned best estimate procedures should be used in the course of fatigue assessment to fulfill e.g. the demands of the KTA code with regard to environmental assisted fatigue. (orig.)

  4. Analysis code for pressure in reactor containment vessel of ATR. CONPOL

    International Nuclear Information System (INIS)

    1997-08-01

    For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

  5. Quality assurance procedures for the CONTAIN severe reactor accident computer code

    International Nuclear Information System (INIS)

    Russell, N.A.; Washington, K.E.; Bergeron, K.D.; Murata, K.K.; Carroll, D.E.; Harris, C.L.

    1991-01-01

    The CONTAIN quality assurance program follows a strict set of procedures designed to ensure the integrity of the code, to avoid errors in the code, and to prolong the life of the code. The code itself is maintained under a code-configuration control system that provides a historical record of changes. All changes are incorporated using an update processor that allows separate identification of improvements made to each successive code version. Code modifications and improvements are formally reviewed and checked. An exhaustive, multilevel test program validates the theory and implementation of all codes changes through assessment calculations that compare the code-predicted results to standard handbooks of idealized test cases. A document trail and archive establish the problems solved by the software, the verification and validation of the software, software changes and subsequent reverification and revalidation, and the tracking of software problems and actions taken to resolve those problems. This document describes in detail the CONTAIN quality assurance procedures. 4 refs., 21 figs., 4 tabs

  6. Development of the containment transient analysis code for the passive reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-05-01

    This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

  7. Evaporation over sump surface in containment studies: code validation on TOSQAN tests

    International Nuclear Information System (INIS)

    Malet, J.; Gelain, T.; Degrees du Lou, O.; Daru, V.

    2011-01-01

    During the course of a severe accident in a Nuclear Power Plant, water can be collected in the sump containment through steam condensation on walls and spray systems activation. The objective of this paper is to present code validation on evaporative sump tests performed on the TOSQAN facility. The ASTEC-CPA code is used as a lumped-parameter code and specific user-defined-functions are developed for the TONUS-CFD code. The tests are air-steam tests, as well as tests with other non-condensable gases (He, CO 2 and SF 6 ) under steady and transient conditions. The results show a good agreement between codes and experiments, indicating a good behaviour of the sump models in both codes. (author)

  8. Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code

    International Nuclear Information System (INIS)

    Perianez Alvarez, V.

    2013-01-01

    The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

  9. Preparation and Elastic Moduli of Germanate Glass Containing Lead and Bismuth

    Directory of Open Access Journals (Sweden)

    Wan M. M. Yunus

    2012-04-01

    Full Text Available This paper reports the rapid melt quenching technique preparation for the new family of bismuth-lead germanate glass (BPG systems in the form of (GeO260–(PbO40−x–(½Bi2O3x where x = 0 to 40 mol%. Their densities with respect of Bi2O3 concentration were determined using Archimedes’ method with acetone as a floatation medium. The current experimental data are compared with those of bismuth lead borate (B2O320–(PbO80−x–(Bi2O3x. The elastic properties of BPG were studied using the ultrasonic pulse-echo technique where both longitudinal and transverse sound wave velocities have been measured in each glass samples at a frequency of 15 MHz and at room temperature. Experimental data shows that all the physical parameters of BPG including density and molar volume, both longitudinal and transverse velocities increase linearly with increasing of Bi2O3 content in the germanate glass network. Their elastic moduli such as longitudinal, shear and Young’s also increase linearly with addition of Bi2O3 but the bulk modulus did not. The Poisson’s ratio and fractal dimensionality are also found to vary linearly with the Bi2O3 concentration.

  10. Preparation and elastic moduli of germanate glass containing lead and bismuth.

    Science.gov (United States)

    Sidek, Hj A A; Bahari, Hamid R; Halimah, Mohamed K; Yunus, Wan M M

    2012-01-01

    This paper reports the rapid melt quenching technique preparation for the new family of bismuth-lead germanate glass (BPG) systems in the form of (GeO(2))(60)-(PbO)(40-) (x)-(½Bi(2)O(3))(x) where x = 0 to 40 mol%. Their densities with respect of Bi(2)O(3) concentration were determined using Archimedes' method with acetone as a floatation medium. The current experimental data are compared with those of bismuth lead borate (B(2)O(3))(20)-(PbO)(80-) (x)-(Bi(2)O(3))(x). The elastic properties of BPG were studied using the ultrasonic pulse-echo technique where both longitudinal and transverse sound wave velocities have been measured in each glass samples at a frequency of 15 MHz and at room temperature. Experimental data shows that all the physical parameters of BPG including density and molar volume, both longitudinal and transverse velocities increase linearly with increasing of Bi(2)O(3) content in the germanate glass network. Their elastic moduli such as longitudinal, shear and Young's also increase linearly with addition of Bi(2)O(3) but the bulk modulus did not. The Poisson's ratio and fractal dimensionality are also found to vary linearly with the Bi(2)O(3) concentration.

  11. Development of fast reactor containment safety analysis code, CONTAIN-LMR. (3) Improvement of sodium-concrete reaction model

    International Nuclear Information System (INIS)

    Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

    2015-01-01

    A computer code, CONTAIN-LMR, is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. Because a sodium-concrete reaction behavior is one of the most important phenomena in the accident, a Sodium-Limestone Concrete Ablation Model (SLAM) has been developed and installed into the original CONTAIN code at Sandia National Laboratories (SNL) in the U.S. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically using a three-region model, containing a pool (sodium and reaction debris) region, a dry (boundary layer (B/L) and dehydrated concrete) region, and a wet (hydrated concrete) region, the application is limited to the reaction between sodium and limestone concrete. In order to apply SLAM to the reaction between sodium and siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved to consider the new chemical reactions between sodium and silicon dioxide. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency (JAEA). It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena. (author)

  12. Simulation of International Standard Problem No. 44 'KAEVER' experiments on aerosol behaviour with the CONTAIN code

    International Nuclear Information System (INIS)

    Kljenak, I.

    2001-01-01

    Experiments on aerosol behavior in a vapor-saturated atmosphere, which were performed in the KAEVER experimental facility and proposed for the OECD International Standard Problem No. 44, were simulated with the CONTAIN thermal-hydraulic computer code. The purpose of the work was to assess the capability of the CONTAIN code to model aerosol condensation and deposition in a containment of a light-water-reactor nuclear power plant at severe accident conditions. Results of dry and wet aerosol concentrations are presented and analyzed.(author)

  13. Analysis of CSNI benchmark test on containment using the code CONTRAN

    International Nuclear Information System (INIS)

    Haware, S.K.; Ghosh, A.K.; Raj, V.V.; Kakodkar, A.

    1994-01-01

    A programme of experimental as well as analytical studies on the behaviour of nuclear reactor containment is being actively pursued. A large number ol' experiments on pressure and temperature transients have been carried out on a one-tenth scale model vapour suppression pool containment experimental facility, simulating the 220 MWe Indian Pressurised Heavy Water Reactors. A programme of development of computer codes is underway to enable prediction of containment behaviour under accident conditions. This includes codes for pressure and temperature transients, hydrogen behaviour, aerosol behaviour etc. As a part of this ongoing work, the code CONTRAN (CONtainment TRansient ANalysis) has been developed for predicting the thermal hydraulic transients in a multicompartment containment. For the assessment of the hydrogen behaviour, the models for hydrogen transportation in a multicompartment configuration and hydrogen combustion have been incorporated in the code CONTRAN. The code also has models for the heat and mass transfer due to condensation and convection heat transfer. The structural heat transfer is modeled using the one-dimensional transient heat conduction equation. Extensive validation exercises have been carried out with the code CONTRAN. The code CONTRAN has been successfully used for the analysis of the benchmark test devised by Committee on the Safety of Nuclear Installations (CSNI) of the Organisation for Economic Cooperation and Development (OECD), to test the numerical accuracy and convergence errors in the computation of mass and energy conservation for the fluid and in the computation of heat conduction in structural walls. The salient features of the code CONTRAN, description of the CSNI benchmark test and a comparison of the CONTRAN predictions with the benchmark test results are presented and discussed in the paper. (author)

  14. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [KHNP-CENTERAL RESEARCH INSTITUTE, Daejeon (Korea, Republic of)

    2013-10-15

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic

  15. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    International Nuclear Information System (INIS)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

    2013-01-01

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic state in

  16. Analysis of LMFBR containment response to an HCDA using a multifield Eulerian code

    International Nuclear Information System (INIS)

    Chu, H.Y.; Chang, Y.W.

    1977-01-01

    This paper describes a computer code, MICE (Multifield Implicit Continuous-fluid Eulerian Containment Code), which is being developed at Argonne National Laboratory (ANL) for the analysis of containment response to a hypothetical core distruptive accident (HCDA). The code is applicable to multifield flow problems where material fields are allowed to have penetrations. Reactor structures are treated as axisymmetrical shells and solved by the large-displacement and small-strain theory. Two sample problems have been performed using the MICE code. The first illustrates the relative motions of the material fields after the initiation of a core disassembly accident. Core support structure and core barrel are modelled as rigid obstacles. The second demonstrates the interactions between fluid and structures. Core expansion and reactor wall deformation at several instants are shown by the computer-generated film plots. (Auth.)

  17. Computing the effects of a contained sodium sheet fire: The 'FEUNA' code

    International Nuclear Information System (INIS)

    Duverger De Cuy, G.

    1979-01-01

    FEUNA is a computer code developed to calculate the thermodynamic effects of a sodium fire in a ventilated or unventilated containment volume. Developed jointly by the CEA/DSN and Novatome, the FEUNA code involves two oxide formation reactions, aerosol generation and deposits, heat transfer by convection, conduction and radiation, gas inflow and outflow through the ventilation system and the relief valves. The code was validated by comparing calculated values with the results of an actual sodium fire in a 400m 3 caisson. (author)

  18. Computing the effects of a contained sodium sheet fire: The 'FEUNA' code

    Energy Technology Data Exchange (ETDEWEB)

    Duverger De Cuy, G [DSN/SESTR, Centre de Cadarache, Saint-Paul-lez-Durance (France)

    1979-03-01

    FEUNA is a computer code developed to calculate the thermodynamic effects of a sodium fire in a ventilated or unventilated containment volume. Developed jointly by the CEA/DSN and Novatome, the FEUNA code involves two oxide formation reactions, aerosol generation and deposits, heat transfer by convection, conduction and radiation, gas inflow and outflow through the ventilation system and the relief valves. The code was validated by comparing calculated values with the results of an actual sodium fire in a 400m{sup 3} caisson. (author)

  19. A directory of computer codes suitable for stress analysis of HLW containers - Compas project

    International Nuclear Information System (INIS)

    1989-01-01

    This document reports the work carried out for the Compas project which looked at the capabilities of various computer codes for the stress analysis of high-level nuclear-waste containers and overpacks. The report concentrates on codes used by the project partners, but also includes a number of the major commercial finite element codes. The report falls into two parts. The first part of the report describes the capabilities of the codes. This includes details of the solution methods used in the codes, the types of analysis which they can carry out and the interfacing with pre - and post - processing packages. This is the more comprehensive section of the report. The second part of the report looks at the performance of a selection of the codes (those used by the project partners). This look at how the codes perform in a number of test problems which require calculations typical of those encountered in the design and analysis of high-level waste containers and overpacks

  20. Development and validation of computer codes for analysis of PHWR containment behaviour

    International Nuclear Information System (INIS)

    Markandeya, S.G.; Haware, S.K.; Ghosh, A.K.; Venkat Raj, V.

    1997-01-01

    In order to ensure that the design intent of the containment of Indian Pressurised Heavy Water Reactors (IPHWRs) is met, both analytical and experimental studies are being pursued at BARC. As a part of analytical studies, computer codes for predicting the behaviour of containment under various accident scenarios are developed/adapted. These include codes for predicting 1) pressure, temperature transients in the containment following either Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB), 2) hydrogen behaviour in respect of its distribution, combustion and the performance of proposed mitigation systems, and 3) behaviour of fission product aerosols in the piping circuits of the primary heat transport system and in the containment. All these codes have undergone thorough validation using data obtained from in-house test facilities or from international sources. Participation in the International Standard Problem (ISP) exercises has also helped in validation of the codes. The present paper briefly describes some of these codes and the various exercises performed for their validation. (author)

  1. Comparison of different LMFBR primary containment codes applied to a Benchmark problem

    International Nuclear Information System (INIS)

    Benuzzi, A.

    1986-01-01

    The Cont Benchmark calculation exercise is a project sponsored by the Containment Loading and Response Group, a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee - CEC. A full-size typical Pool type LMFBR undergoing a postulated Core Disruptive Accident (CDA) has been defined by Belgonucleaire-Brussels under a study contract financed by the CEC and has been submitted to seven containment code calculations. The results of these calculations are presented and discussed in this paper

  2. The modelling of wall condensation with noncondensable gases for the containment codes

    Energy Technology Data Exchange (ETDEWEB)

    Leduc, C.; Coste, P.; Barthel, V.; Deslandes, H. [Commissariat a l`Energi Atomique, Grenoble (France)

    1995-09-01

    This paper presents several approaches in the modelling of wall condensation in the presence of noncondensable gases for containment codes. The lumped-parameter modelling and the local modelling by 3-D codes are discussed. Containment analysis codes should be able to predict the spatial distributions of steam, air, and hydrogen as well as the efficiency of cooling by wall condensation in both natural convection and forced convection situations. 3-D calculations with a turbulent diffusion modelling are necessary since the diffusion controls the local condensation whereas the wall condensation may redistribute the air and hydrogen mass in the containment. A fine mesh modelling of film condensation in forced convection has been in the developed taking into account the influence of the suction velocity at the liquid-gas interface. It is associated with the 3-D model of the TRIO code for the gas mixture where a k-{xi} turbulence model is used. The predictions are compared to the Huhtiniemi`s experimental data. The modelling of condensation in natural convection or mixed convection is more complex. As no universal velocity and temperature profile exist for such boundary layers, a very fine nodalization is necessary. More simple models integrate equations over the boundary layer thickness, using the heat and mass transfer analogy. The model predictions are compared with a MIT experiment. For the containment compartments a two node model is proposed using the lumped parameter approach. Heat and mass transfer coefficients are tested on separate effect tests and containment experiments. The CATHARE code has been adapted to perform such calculations and shows a reasonable agreement with data.

  3. TRAC-CFD code integration and its application to containment analysis

    International Nuclear Information System (INIS)

    Tahara, M.; Arai, K.; Oikawa, H.

    2004-01-01

    Several safety systems utilizing natural driving force have been recently adopted for operating reactors, or applied to next-generation reactor design. Examples of these safety systems are the Passive Containment Cooling System (PCCS) and the Drywell Cooler (DWC) for removing decay heat, and the Passive Auto-catalytic Recombiner (PAR) for removing flammable gas in reactor containment during an accident. DWC is used in almost all Boiling Water Reactors (BWR) in service. PAR has been introduced for some reactors in Europe and will be introduced for Japanese reactors. PCCS is a safety device of next-generation BWR. The functional mechanism of these safety systems is closely related to the transient of the thermal-hydraulic condition of the containment atmosphere. The performance depends on the containment atmospheric condition, which is eventually affected by the mass and energy changes caused by the safety system. Therefore, the thermal fluid dynamics in the containment vessel should be appropriately considered in detail to properly estimate the performance of these systems. A computational fluid dynamics (CFD) code is useful for evaluating detailed thermal hydraulic behavior related to this equipment. However, it also requires a considerable amount of computational resources when it is applied to whole containment system transient analysis. The paper describes the method and structure of the integrated analysis tool, and discusses the results of its application to the start-up behavior analysis of a containment cooling system, a drywell local cooler. The integrated analysis code was developed and applied to estimate the DWC performance during a severe accident. The integrated analysis tool is composed of three codes, TRAC-PCV, CFD-DW and TRAC-CC, and analyzes the interaction of the natural convection and steam condensation of the DWC as well as analyzing the thermal hydraulic transient behavior of the containment vessel during a severe accident in detail. The

  4. Review and assessment of thermodynamic and transport properties for the CONTAIN Code

    International Nuclear Information System (INIS)

    Valdez, G.D.

    1988-12-01

    A study was carried out to review available data and correlations on the thermodynamic and transport properties of materials applicable to the CONTAIN computer code. CONTAIN is the NRC's best-estimate, mechanistic computer code for modeling containment response to a severe accident. Where appropriate, recommendations have been made for suitable approximations for material properties of interests. Based on a modified Benedict-Webb-Rubin (BWR) equation of state, a procedure is introduced for calculating thermodynamic properties for common gases in the CONTAIN code. These gases are nitrogen, oxygen, hydrogen, carbon dioxide, carbon monoxide, steam, helium, and argon. The thermodynamic equations for density, currently represented in CONTAIN by relatively simple fits, were independently checked and are recommended to be replaced by the Lee-Kesler equation of state which substantially improves accuracy without too much sacrifice in computational efficiency. The accuracy of the calculated values have been found to be generally acceptable. Various correlations and models for single component gas transport properties, viscosity and thermal conductivity, were also assessed with available experimental data. When a suitable correlation or model was not available, transport properties were obtained by performing least-squares fit on experimental data. 50 refs., 126 figs., 3 tabs

  5. The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

    International Nuclear Information System (INIS)

    Wang Jianyu; Zhang Shenru; Min Yuanyou

    1994-01-01

    To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

  6. Simulation of atmosphere stratification in the HDR test facility with the CONTAIN code

    International Nuclear Information System (INIS)

    Skerlavaj, A.; Mavko, B.; Kljenak, I.

    2001-01-01

    The test E11.2 'Hydrogen distribution in loop flow geometry', which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.(author)

  7. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    International Nuclear Information System (INIS)

    Raske, D.T.; Wang, Z.

    1992-01-01

    The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

  8. Large scale fire experiments in the HDR containment as a basis for fire code development

    International Nuclear Information System (INIS)

    Hosser, D.; Dobbernack, R.

    1993-01-01

    Between 1984 and 1991 7 different series of large scale fire experiments and related numerical and theoretical investigations have been performed in the containment of a high pressure reactor in Germany (known as HDR plant). The experimental part included: gas burner tests for checking the containment behaviour; naturally ventilated fires with wood cribs; naturally and forced ventilated oil pool fires; naturally and forced ventilated cable fires. Many results of the oil pool and cable fires can directly be applied to predict the impact of real fires at different locations in a containment on mechanical or structural components as well as on plant personnel. But the main advantage of the measurements and observations was to serve as a basis for fire code development and validation. Different types of fire codes have been used to predict in advance or evaluate afterwards the test results: zone models for single room and multiple room configurations; system codes for multiple room configurations; field models for complex single room configurations. Finally, there exist codes of varying degree of specialization which have proven their power and sufficient exactness to predict fire effects as a basis for optimum fire protection design. (author)

  9. German Studies in America. German Studies Notes.

    Science.gov (United States)

    Sander, Volkmar; Osterle, Heinz D.

    This volume contains two papers, "German Studies in America," by Volkmar Sander, and "Historicism, Marxism, Structuralism: Ideas for German Culture Courses," by Heinz D. Osterle. The first paper discusses the position of German studies in the United States today. The greatest challenge comes from low enrollments; therefore,…

  10. Object-Oriented Programming in the Development of Containment Analysis Code

    International Nuclear Information System (INIS)

    Han, Tae Young; Hong, Soon Joon; Hwang, Su Hyun; Lee, Byung Chul; Byun, Choong Sup

    2009-01-01

    After the mid 1980s, the new programming concept, Object-Oriented Programming (OOP), was introduced and designed, which has the features such as the information hiding, encapsulation, modularity and inheritance. These offered much more convenient programming paradigm to code developers. The OOP concept was readily developed into the programming language as like C++ in the 1990s and is being widely used in the modern software industry. In this paper, we show that the OOP concept is successfully applicable to the development of safety analysis code for containment and propose the more explicit and easy OOP design for developers

  11. Assessment of GOTHIC and TRACE codes against selected PANDA experiments on a Passive Containment Condenser

    Energy Technology Data Exchange (ETDEWEB)

    Papini, Davide, E-mail: davide.papini@psi.ch; Adamsson, Carl; Andreani, Michele; Prasser, Horst-Michael

    2014-10-15

    Highlights: • Code comparison on the performance of a Passive Containment Condenser. • Simulation of separate effect tests with pure steam and non-condensable gases. • Role of the secondary side and accuracy of pool boiling models are discussed. • GOTHIC and TRACE predict the experimental performance with slight underestimation. • Recirculatory flow pattern with injection of light non-condensable gas is inferred. - Abstract: Typical passive safety systems for ALWRs (Advanced Light Water Reactors) rely on the condensation of steam to remove the decay heat from the core or the containment. In the present paper the three-dimensional containment code GOTHIC and the one-dimensional system code TRACE are compared on the calculation of a variety of phenomena characterizing the response of a passive condenser submerged in a boiling pool. The investigation addresses the conditions of interest for the Passive Containment Cooling System (PCCS) proposed for the ESBWR (Economic Simplified Boiling Water Reactor). The analysis of selected separate effect tests carried out on a PCC (Passive Containment Condenser) unit in the PANDA large-scale thermal-hydraulic facility is presented to assess the code predictions. Both pure steam conditions (operating pressure of 3 bar, 6 bar and 9 bar) and the effect on the condensation heat transfer of non-condensable gases heavier than steam (air) and lighter than steam (helium) are considered. The role of the secondary side (pool side) heat transfer on the condenser performance is examined too. In general, this study shows that both the GOTHIC and TRACE codes are able to reasonably predict the heat transfer capability of the PCC as well as the influence of non-condensable gas on the system. A slight underestimation of the condenser performance is obtained with both codes. For those tests where the experimental and simulated efficiencies agree better the possibility of compensating errors among different parts of the heat transfer

  12. Comparison of aerosol behavior codes with experimental results from a sodium fire in a containment

    International Nuclear Information System (INIS)

    Lhiaubet, G.; Kissane, M.P.; Seino, H.; Miyake, O.; Himeno, Y.

    1990-01-01

    The containment expert group (CONT), a subgroup of the CEC fast reactor Safety Working Group (SWG), has carried out several studies on the behavior of sodium aerosols which might form in a severe fast reactor accident during which primary sodium leaks into the secondary containment. These studies comprise an intercalibration of measurement devices used to determine the aerosol particle size spectrum, and the analysis and comparison of codes applied to the determination of aerosol behavior in a reactor containment. The paper outlines the results of measurements of typical data made for aerosols produced in a sodium fire and their comparison with results from different codes (PARDISEKO, AEROSIM, CONTAIN, AEROSOLS/B2). The sodium fire experiment took place at CEN-Cadarache (France) in a 400 m 3 vessel. The fire lasted 90 minutes and the aerosol measurements were made over 10 hours at different locations inside the vessel. The results showed that the suspended mass calculated along the time with different codes was in good agreement with the experiment. However, the calculated aerosol deposition on the walls was diverging and always significantly lower than the measured values

  13. Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

    International Nuclear Information System (INIS)

    Papini, Davide; Grgic, Davor; Cammi, Antonio; Ricotti, Marco E.

    2011-01-01

    Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes). The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.

  14. An estimation of uncertainties in containment P/T analysis using CONTEMPT/LT code

    International Nuclear Information System (INIS)

    Kang, Y.M.; Park, G.C.; Lee, U.C.; Kang, C.S.

    1991-01-01

    In a nuclear power plant, the containment design pressure and temperature (P/T) have been established based on the unrealistic conservatism with suffering from a drawback in the economics. Thus, it is necessary that the uncertainties of design P/T values have to be well defined through an extensive uncertainty analysis with plant-specific input data and or models used in the computer code. This study is to estimate plant-specific uncertainties of containment design P/T using the Monte Carlo method in Kori-3 reactor. Kori-3 plant parameters and Uchida heat transfer coefficient are selected to be treated statistically after the sensitivity study. The Monte Carlo analysis has performed based on the response surface method with the CONTEMPT/LT code and Latin Hypercube sampling technique. Finally, the design values based on 95 %/95 % probability are compared with worst estimated values to assess the design margin. (author)

  15. Validation of computer code TRAFIC used for estimation of charcoal heatup in containment ventilation systems

    International Nuclear Information System (INIS)

    Yadav, D.H.; Datta, D.; Malhotra, P.K.; Ghadge, S.G.; Bajaj, S.S.

    2005-01-01

    Full text of publication follows: Standard Indian PHWRs are provided with a Primary Containment Filtration and Pump-Back System (PCFPB) incorporating charcoal filters in the ventilation circuit to remove radioactive iodine that may be released from reactor core into the containment during LOCA+ECCS failure which is a Design Basis Accident for containment of radioactive release. This system is provided with two identical air circulation loops, each having 2 full capacity fans (1 operating and 1 standby) for a bank of four combined charcoal and High Efficiency Particulate Activity (HEPA) filters, in addition to other filters. While the filtration circuit is designed to operate under forced flow conditions, it is of interest to understand the performance of the charcoal filters, in the event of failure of the fans after operating for some time, i.e., when radio-iodine inventory is at its peak value. It is of interest to check whether the buoyancy driven natural circulation occurring in the filtration circuit is sufficient enough to keep the temperature in the charcoal under safe limits. A computer code TRAFIC (Transient Analysis of Filters in Containment) was developed using conservative one dimensional model to analyze the system. Suitable parametric studies were carried out to understand the problem and to identify the safety of existing system. TRAFIC Code has two important components. The first one estimates the heat generation in charcoal filter based on 'Source Term'; while the other one performs thermal-hydraulic computations. In an attempt validate the Code, experimental studies have been carried out. For this purpose, an experimental set up comprising of scaled down model of filtration circuit with heating coils embedded in charcoal for simulating the heating effect due to radio iodine has been constructed. The present work of validation consists of utilizing the results obtained from experiments conducted for different heat loads, elevations and adsorbent

  16. Corrections and additions to CONTEMPT-LT computer codes for containment analysis

    International Nuclear Information System (INIS)

    Eerikaeinen, Lauri.

    1980-01-01

    The report presents a new version of CONTEMPT-LT/26 tainment code. The corrections and additions are applicable also to other CONTEMPT-LT versions. Thermodynamical background of corrections are shortly described, and in addition, some essential points which should be taken into account in the analysis of a pressure suppression containment have been pointed out. The results obtained by the corrected version have been compared to those calculated by the original program, and to the measured data in the Marviken containment experiment No 10. Finally, it has been indicated that for reliable pressure suppression analysis a wide ranging condensation model for air-steam mixture is necessary. (author)

  17. JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.

    1983-12-01

    The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

  18. Structural dynamics in LMFBR containment analysis. A brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper gives a brief survey of the computational methods and codes available for LMFBR containment analysis. The various numerical methods commonly used in the computer codes are compared. It provides the reactor engineers to up-to-date information on the development of structural dynamics in LMFBR containment analysis. It can also be used as a basis for the selection of the numerical method in the future code development. First, the commonly used finite-difference expressions in the Lagrangian codes will be compared. Sample calculations will be used as a basis for discussing and comparing the accuracy of the various finite-difference representations. The distortion of the meshes will also be compared; the techniques used for eliminating the numerical instabilities will be discussed and compared using examples. Next, the numerical methods used in the Eulerian formulation will be compared, first among themselves and then with the Lagrangian formulations. Special emphasis is placed on the effect of mass diffusion of the Eulerian calculation on the propagation of discontinuities. Implicit and explicit numerical integrations will be discussed and results obtained from these two techniques will be compared. Then, the finite-element methods are compared with the finite-difference methods. The advantages and disadvantages of the two methods will be discussed in detail, together with the versatility and ease of application of the method to containment analysis having complex geometries. It will also be shown that the finite-element equations for a constant-pressure fluid element is identical to the finite-difference equations using contour integrations. Finally, conclusions based on this study will be given

  19. Modelling of water sump evaporation in a CFD code for nuclear containment studies

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Bessiron, M., E-mail: matthieu.bessiron@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Perrotin, C., E-mail: christophe.perrotin@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France)

    2011-05-15

    Highlights: We model sump evaporation in the reactor containment for CFD codes. The sump is modelled by an interface temperature and an evaporation mass flow-rate. These two variables are modelled using energy and mass balance. Results are compared with specific experiments in a 7 m3 vessel (Tonus Qualification ANalytique, TOSQAN). A good agreement is observed, for pressure, temperatures, mass flow-rates. - Abstract: During the course of a hypothetical severe accident in a pressurized water reactor (PWR), water can be collected in the sump containment through steam condensation on walls and spray systems activation. This water is generally under evaporation conditions. The objective of this paper is twofold: to present a sump model developed using external user-defined functions for the TONUS-CFD code and to perform a first detailed comparison of the model results with experimental data. The sump model proposed here is based on energy and mass balance and leads to a good agreement between the numerical and the experimental results. Such a model can be rather easily added to any CFD code for which boundary conditions, such as injection temperature and mass flow-rate, can be modified by external user-defined functions, depending on the atmosphere conditions.

  20. Modernization of the graphics post-processors of the Hamburg German Climate Computer Center Carbon Cycle Codes

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, E.J.; McNeilly, G.S.

    1994-03-01

    The existing National Center for Atmospheric Research (NCAR) code in the Hamburg Oceanic Carbon Cycle Circulation Model and the Hamburg Large-Scale Geostrophic Ocean General Circulation Model was modernized and reduced in size while still producing an equivalent end result. A reduction in the size of the existing code from more than 50,000 lines to approximately 7,500 lines in the new code has made the new code much easier to maintain. The existing code in Hamburg model uses legacy NCAR (including even emulated CALCOMP subrountines) graphics to display graphical output. The new code uses only current (version 3.1) NCAR subrountines.

  1. CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

    1993-01-01

    The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

  2. Nupack, the new ASME code for radioactive material transportation packaging containments

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as Nupack, has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used for the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper

  3. Advanced local dose rate calculations with the Monte Carlo code MCNP for plutonium nitrate storage containers

    International Nuclear Information System (INIS)

    Quade, U.

    1994-01-01

    Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de

  4. CODE ACCEPTANCE OF A NEW JOINING TECHNOLOGY FOR STORAGE CONTAINMENTS (REISSUE)

    International Nuclear Information System (INIS)

    Cannel, G.R.; Grant, G.J.; Hill, B.E.

    2009-01-01

    One of the activities associated with cleanup throughout the Department of Energy (DOE) complex is packaging radioactive materials into storage containers. Much of this work will be performed in high-radiation environments requiring fully remote operations, for which existing, proven systems do not currently exist. These conditions require a process that is capable of producing acceptable (defect-free) welds on a consistent basis; the need to perform weld repair, under fully-remote operations, can be extremely costly and time consuming. Current closure-welding technologies (fusion welding) are not well suited for this application and will present risk to cleanup cost and schedule. To address this risk, Fluor and the Pacific Northwest National Laboratory (PNNL) are proposing that a new and emerging joining technology, Friction Stir Welding (FSW), be considered for this work. FSW technology has been demonstrated in other industries (aerospace and marine) to produce near flaw-free welds on a consistent basis. FSW is judged capable of providing the needed performance for fully-remote closure welding of containers for radioactive materials for the following reasons: FSW is a solid-state process; material is not melted. FSW does not produce the type of defects associated with fusion welding, e.g., solidification-induced porosity, cracking, and distortion due to weld shrinkage. In addition, because FSW is a low-heat input process, material properties (mechanical, corrosion and environmental) experience less degradation in the heat affected zones than do fusion welds. When compared to fusion processes, FSW produces extremely high weld quality. FSW is performed using machine-tool technology. The equipment is simple and robust and well-suited for high radiation, fully-remote operations compared to the relatively complex equipment associated with fusion-welding processes. Additionally, for standard wall thicknesses of radioactive materials containers, the FSW process can

  5. Container-code recognition system based on computer vision and deep neural networks

    Science.gov (United States)

    Liu, Yi; Li, Tianjian; Jiang, Li; Liang, Xiaoyao

    2018-04-01

    Automatic container-code recognition system becomes a crucial requirement for ship transportation industry in recent years. In this paper, an automatic container-code recognition system based on computer vision and deep neural networks is proposed. The system consists of two modules, detection module and recognition module. The detection module applies both algorithms based on computer vision and neural networks, and generates a better detection result through combination to avoid the drawbacks of the two methods. The combined detection results are also collected for online training of the neural networks. The recognition module exploits both character segmentation and end-to-end recognition, and outputs the recognition result which passes the verification. When the recognition module generates false recognition, the result will be corrected and collected for online training of the end-to-end recognition sub-module. By combining several algorithms, the system is able to deal with more situations, and the online training mechanism can improve the performance of the neural networks at runtime. The proposed system is able to achieve 93% of overall recognition accuracy.

  6. Sodium spray and jet fire model development within the CONTAIN-LMR code

    International Nuclear Information System (INIS)

    Scholtyssek, W.

    1993-01-01

    An assessment was made of the sodium spray fire model implemented in the CONTAIN code. The original droplet burn model, which was based on the NACOM code, was improved in several aspects, especially concerning evaluation of the droplet burning rate, reaction chemistry and heat balance, spray geometry and droplet motion, and consistency with CONTAIN standards of gas property evaluation. An additional droplet burning model based on a proposal by Krolikowski was made available to include the effect of the chemical equilibrium conditions at the flame temperature. The models were validated against single-droplet burn experiments as well as spray and jet fire experiments. Reasonable agreement was found between the two burn models and experimental data. When the gas temperature in the burning compartment reaches high values, the Krolikowski model seems to be preferable. Critical parameters for spray fire evaluation were found to be the spray characterization, especially the droplet size, which largely determines the burning efficiency, and heat transfer conditions at the interface between the atmosphere and structures, which controls the thermal hydraulic behavior in the burn compartment

  7. Analysis of LMFBR containment response to an HCDA using a multifield Eulerian code

    International Nuclear Information System (INIS)

    Chu, H.Y.; Chang, Y.W.

    1977-01-01

    During a hypothetical core disruptive accident (HCDA), a core meltdown may cause the fuel cladding to rupture and the fuel fragments to penetrate into the sodium coolant. The heat in the molten fuel may cause the liquid sodium to boil, changing its phase. The interactions between materials are so complicated that a single-material model with homogenized material properties is not adequate. In order to analyze the above phenomena more realistically, a Multifield Implicit Continuous-Fluid Eulerian containment code (MICE) is being developed at Argonne National Laboratory (ANL) to solve the multifield fluid-flow problems in which the interpenetrations of materials, heat transfer, and phase changes are considered in the analysis. The hydrodynamics of the MICE code is based upon the implicit multifield (IMF) method developed by Harlow and Amsden. A partial donor-cell formulation is used for the calculation of the convective fluxes to minimize the truncation errors, while the Newton-Raphson method is used for the numerical iterations. An implicit treatment of the mass convection together with the equation of state for each material enables the method to be applicable to both compressible and incompressible flows. A partial implicit treatment of the momentum-exchange functions allows the coupling drag forces between two material fields to range from very weak to those strong enough to tie the fields completely. The differential equations and exchange functions used in the MICE code, and the treatment of the fluid and structure interactions as well as the numerical procedure are described. Two sample calculations are given to illustrate the present capability of the MICE code

  8. Summary of aerosol code-comparison results for LWR aerosol containment tests LA1, LA2, and LA3

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1987-01-01

    The light-water reactor (LWR) aerosol containment experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities for the LACE tests are being coordinated at the Oak Ridge National Laboratory. For each of the six experiments, pretest calculations (for code-to-code comparisons) and blind post-test calculations (for code-to-test data comparisons) are being performed. This paper presents a summary of the pretest aerosol-code results for tests LA1, LA2, and LA3

  9. Analysis of the Phebus FPT0 containment thermal hydraulics with the Jericho and Trio-VF codes

    International Nuclear Information System (INIS)

    Layly, V.D.; Spitz, P.; Mailliat, A.

    1994-01-01

    This paper presents the analysis of the thermal hydraulic behavior of the containment, during the Phebus FPT0 test performed on December 2, 1993, with the Jericho code which deals with the thermal hydraulics of containment in the severe accident field. This code is part of Escadre which is the French system of codes in charge of predicting PWR severe accidents. After summarizing the relevant Jericho code characteristics and the preliminary assessment work for the Phebus conditions, we briefly describe the REPF 502 test facility and report the thermal hydraulic FPT0 experimental protocol. Then, the experiment / Jericho calculation comparisons are analysed. Because the Jericho code assumes a well-mixed atmosphere, some additional 3-D calculations have been carried out in order to get further insight on the convection flow patterns and qualify the well-mixed atmosphere assumption in the Phebus containment. (author). 9 refs., 12 figs

  10. Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

  11. Development of Evaluation Technology for Hydrogen Combustion in containment and Accident Management Code for CANDU

    International Nuclear Information System (INIS)

    Kim, S. B.; Kim, D. H.; Song, Y. M.

    2011-08-01

    For a licensing of nuclear power plant(NPP) construction and operation, the hydrogen combustion and hydrogen mitigation system in the containment is one of the important safety issues. Hydrogen safety and its control for the new NPPs(Shin-Wolsong 1 and 2, Shin-Ulchin 1 and 2) have been evaluated in detail by using the 3-dimensional analysis code GASFLOW. The experimental and computational studies on the hydrogen combustion, and participations of the OEDE/NEA programs such as THAI and ISP-49 secures the resolving capabilities of the hydrogen safety and its control for the domestic nuclear power plants. ISAAC4.0, which has been developed for the assessment of severe accident management at CANDU plants, was already delivered to the regulatory body (KINS) for the assessment of the severe accident management guidelines (SAMG) for Wolsong units 1 to 4, which are scheduled to be submitted to KINS. The models for severe accident management strategy were newly added and the graphic simulator, CAVIAR, was coupled to addition, the ISAAC computer code is anticipated as a platform for the development and maintenance of Wolsong plant risk monitor and Wolsong-specific SAMG

  12. Nupack, the new Asme code for radioactive material transportation packaging containments

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as 'Nupack', has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper. Participation in the Nupack development work described in this paper was supported by the U.S. Department of Energy. (authors)

  13. A simplified model of Passive Containment Cooling System in a CFD code

    International Nuclear Information System (INIS)

    Jiang, X.W.; Studer, E.; Kudriakov, S.

    2013-01-01

    Highlights: ► We have built a condensing model using Navier–Stokes equations in CAST3M code. ► We have done a benchmark work on the condensing model using the COPAIN tests data. ► We have built an evaporating model according to Aiello's model in CAST3M code. ► We used Kang and Park's film evaporation tests data to validate the model. ► An integrated model was derived by coupling two individual models with a steel plate. -- Abstract: In this paper, we built up a simplified model of the Passive Containment Cooling System in a CFD code, including a steel plate, a condensing channel and an evaporating channel. In the inner side of the plate, the condensing channel is supposed to be the source of heat transfer into the steel plate. Along the outer side, an evaporating falling film is used to extract the heat from the steel plate. Upward flow of air is also considered along the evaporating film. In the condensing channel, a flow solver based on an asymptotic model of the Navier–Stokes equations at the low Mach number regime and two turbulence models (Buleev's model and Chien's k–ε model) are considered. The condensing channel model was used to model the COPAIN test, the computed heat flux and condensation rate were compared with the experimental data. In the evaporating channel, a simplified model developed by Aiello and Ciofalo (2009) was used, which considered the heat and mass balance between the falling film and the ascending air flow. The model was validated for two cases: a dry wall case and a completely wet wall case. In the former case, the results were compared with 2D predictions obtained by using the CFX-4 CFD code. In the latter case, the results were compared with experimental data obtained by Kang and Park. The comparison showed a satisfactory agreement on heat transfer rates, despite some overprediction depending on the air velocity. At the end, the condensing channel model and the evaporating channel model were coupled by the steel plate

  14. Assessment of Prediction Capabilities of COCOSYS and CFX Code for Simplified Containment

    Directory of Open Access Journals (Sweden)

    Jia Zhu

    2016-01-01

    Full Text Available The acceptable accuracy for simulation of severe accident scenarios in containments of nuclear power plants is required to investigate the consequences of severe accidents and effectiveness of potential counter measures. For this purpose, the actual capability of CFX tool and COCOSYS code is assessed in prototypical geometries for simplified physical process-plume (due to a heat source under adiabatic and convection boundary condition, respectively. Results of the comparison under adiabatic boundary condition show that good agreement is obtained among the analytical solution, COCOSYS prediction, and CFX prediction for zone temperature. The general trend of the temperature distribution along the vertical direction predicted by COCOSYS agrees with the CFX prediction except in dome, and this phenomenon is predicted well by CFX and failed to be reproduced by COCOSYS. Both COCOSYS and CFX indicate that there is no temperature stratification inside dome. CFX prediction shows that temperature stratification area occurs beneath the dome and away from the heat source. Temperature stratification area under adiabatic boundary condition is bigger than that under convection boundary condition. The results indicate that the average temperature inside containment predicted with COCOSYS model is overestimated under adiabatic boundary condition, while it is underestimated under convection boundary condition compared to CFX prediction.

  15. Containment PT Analysis in case of installation of PCCS using CAP Code

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Jun; Hong, Soon Joon [Future and Challenge Technology Co., Yongin (Korea, Republic of); Kim, Gon Han; Cheon, Jong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The goal of this project is developing of the conceptual design of containment passive cooling system, including PMCCS (Passive Molten Core Cooling System) and furthermore, applying to APR+ and iPOWER reactor finally. Advanced researches on PCCS were carried out by global nuclear industrial companies and the several types of reactor, such as AP1000, ESBWR and VVER, were introduced. Using the different way, these reactor types, however, are devised to mitigate the consequence of LOCA accident and remove the long-term decay heat. To confirm the reliability of PCCS function during an accident progress, experimental proofs and computational evaluations are vital. To evaluate the PCCS performance, CAP code simulation is preliminarily conducted on the ShinKori 3/4 DEDLSB accident. Two condensation models, Uchida and Dehbi's correlation, are tested for the condensation model applied on PCCS condensation tube outside surface. As compared with the existing spray system, it revealed the good performance in terms of containment pressure reduction. On the other hand, re-pressurization with the start of PCCT coolant temperature increment is observed also.

  16. Assessment of cavity dispersal correlations for possible implementation in the CONTAIN code

    International Nuclear Information System (INIS)

    Williams, D.C.; Griffith, R.O.

    1996-02-01

    Candidate models and correlations describing entrainment and dispersal of core debris from reactor cavities in direct containment heating (DCH) event, are assessed against a data base of approximately 600 experiments performed previously at Brookhaven National Laboratory and Sandia National Laboratories reactor cavities was studied. Cavity geometries studied are those of the Surry and Zion nuclear power plants and scale factors of 1/42 and 1/10 were studied for both geometries. Other parameters varied in the experiments include gas pressure driving the dispersal, identities of the driving gas and of the simulant fluid, orifice diameter in the pressure vessel, and volume of the gas pressure vessel. Correlations were assessed in terms of their ability to reproduce the observed trends in the fractions dispersed as the experimental parameters were varied. For the fraction of the debris dispersed, the correlations recommended for inclusion in the CONTAIN code are the Tutu-Ginsberg correlations, the integral form of the correlation proposed by Levy and a modified form of the Whalley-Hewitt correlation. For entrainment rates, the recommended correlations are the time-dependent forms of the Levy correlation, a correlation suggested by Tutu, and the modified Whalley-Hewitt correlation

  17. ZOCO V - a computer code for the calculation of time-dependent spatial pressure distribution in reactor containments

    International Nuclear Information System (INIS)

    Mansfeld, G.; Schally, P.

    1978-06-01

    ZOCO V is a computer code which can calculate the time- and space- dependent pressure distribution in containments of water-cooled nuclear power reactors (both full pressure containments and pressure suppression systems) following a loss-of-coolant accident, caused by the rupture of a main coolant or steam pipe

  18. [Placement of children and adolescents following seclusion and restraint actions–a study on family-court approvals of minors in youth welfare, child and adolescent psychiatry and jail according to Para. 1631 German Civil Code].

    Science.gov (United States)

    Kölch, Michael; Vogel, Harald

    2016-01-01

    According to German law (Para. 1631b German Civil Code), the placement of children and adolescents following seclusion and restraint actions must be approved by a family court. We analyzed the family court data of a court district in Berlin (Tempelhof-Kreuzberg) concerning cases of “placement of minors” between 2008 and 2011. A total of 474 such procedures were discovered. After data clearing and correction of cases (e. g., because of emergency interventions of the youth welfare system taking children into custody according to Para. 42, German Civil Code VIII), 376 cases remained. Of these 376 procedures in the years 2008 to 2011, 127 cases concerned children and adolescents according to Para. 1631b German Civil Code, and 249 procedures were settled either by dismissal, withdrawal or by repealing the initial decision to place the child with restrain or seclusion by means of an interim order or by filing an appeal against the final decision. Of the 127 procedures, 68 concerned girls, who were on average slightly younger than boys (14.5 years vs. 15.1 years). In two thirds of the procedures, the children and adolescents were German citizens. The majority of youths involved were living at home at the time of the procedure, but in 15 % of the case the youths were homeless. Most of the adolescents were treated with restraint in child and adolescent psychiatry. The most frequently quoted reasons for seclusion were substance abuse, suicide risk and running away from home/being homeless.

  19. Container for waste, identification code reading device thereof, method and system for controlling waste by using them

    International Nuclear Information System (INIS)

    Kikuchi, Takashi; Yoshida, Tomiji; Omote, Tatsuyuki.

    1991-01-01

    In the conventional method of controlling waste containers by labels attached thereto, the data relevant to wastes contained in the waste containers are limited. Further, if the label should be peeled off, there is a possibility that the wastes therein can no more be identified. Then, in the present invention, an identification plate is previously attached, to which mechanically readable codes or visually readable letters or numerical figures are written. Then, the identification codes can be read in a remote control manner at high speed and high reliability and the waste containers can be managed only by the identification codes of the containers. Further, the identification codes on the container are made so as to be free from aging degradation, thereby enabling to manage waste containers for long time storage. With such a constitution, since data can be inputted from an input terminal and a great amount of data such as concerning the source of wastes can be managed collectively on a software, the data can be managed easily. (T.M.)

  20. A research on verification of the models in the CONTAIN Code and the uncertainty reduction method for containment integrity evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Moo Hwan; Seo, Kyoung Woo [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    2000-03-15

    The final goal of this research is to evaluate synthetic results of DCH issue and expose accurate methodology to assess containment integrity about operating PWR in Korea. This research is aimed to expose methodology for synthetic resolution of the DCH issue for KSNPP, and make the guide of DCH issue for containment integrity which will be used to design to nuclear power plants.

  1. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    International Nuclear Information System (INIS)

    Kalkahoran, Omid Noori; Ahangari, Rohollah; Shirani, Amir Saied

    2016-01-01

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results

  2. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Kalkahoran, Omid Noori; Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

  3. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

  4. 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)

  5. International standard problem (ISP) no. 41 follow up exercise: Containment iodine computer code exercise: parametric studies

    Energy Technology Data Exchange (ETDEWEB)

    Ball, J.; Glowa, G.; Wren, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ewig, F. [GRS Koln (Germany); Dickenson, S. [AEAT, (United Kingdom); Billarand, Y.; Cantrel, L. [IPSN (France); Rydl, A. [NRIR (Czech Republic); Royen, J. [OECD/NEA (France)

    2001-11-01

    This report describes the results of the second phase of International Standard Problem (ISP) 41, an iodine behaviour code comparison exercise. The first phase of the study, which was based on a simple Radioiodine Test Facility (RTF) experiment, demonstrated that all of the iodine behaviour codes had the capability to reproduce iodine behaviour for a narrow range of conditions (single temperature, no organic impurities, controlled pH steps). The current phase, a parametric study, was designed to evaluate the sensitivity of iodine behaviour codes to boundary conditions such as pH, dose rate, temperature and initial I{sup -} concentration. The codes used in this exercise were IODE(IPSN), IODE(NRIR), IMPAIR(GRS), INSPECT(AEAT), IMOD(AECL) and LIRIC(AECL). The parametric study described in this report identified several areas of discrepancy between the various codes. In general, the codes agree regarding qualitative trends, but their predictions regarding the actual amount of volatile iodine varied considerably. The largest source of the discrepancies between code predictions appears to be their different approaches to modelling the formation and destruction of organic iodides. A recommendation arising from this exercise is that an additional code comparison exercise be performed on organic iodide formation, against data obtained front intermediate-scale studies (two RTF (AECL, Canada) and two CAIMAN facility, (IPSN, France) experiments have been chosen). This comparison will allow each of the code users to realistically evaluate and improve the organic iodide behaviour sub-models within their codes. (author)

  6. International standard problem (ISP) no. 41 follow up exercise: Containment iodine computer code exercise: parametric studies

    International Nuclear Information System (INIS)

    Ball, J.; Glowa, G.; Wren, J.; Ewig, F.; Dickenson, S.; Billarand, Y.; Cantrel, L.; Rydl, A.; Royen, J.

    2001-11-01

    This report describes the results of the second phase of International Standard Problem (ISP) 41, an iodine behaviour code comparison exercise. The first phase of the study, which was based on a simple Radioiodine Test Facility (RTF) experiment, demonstrated that all of the iodine behaviour codes had the capability to reproduce iodine behaviour for a narrow range of conditions (single temperature, no organic impurities, controlled pH steps). The current phase, a parametric study, was designed to evaluate the sensitivity of iodine behaviour codes to boundary conditions such as pH, dose rate, temperature and initial I - concentration. The codes used in this exercise were IODE(IPSN), IODE(NRIR), IMPAIR(GRS), INSPECT(AEAT), IMOD(AECL) and LIRIC(AECL). The parametric study described in this report identified several areas of discrepancy between the various codes. In general, the codes agree regarding qualitative trends, but their predictions regarding the actual amount of volatile iodine varied considerably. The largest source of the discrepancies between code predictions appears to be their different approaches to modelling the formation and destruction of organic iodides. A recommendation arising from this exercise is that an additional code comparison exercise be performed on organic iodide formation, against data obtained front intermediate-scale studies (two RTF (AECL, Canada) and two CAIMAN facility, (IPSN, France) experiments have been chosen). This comparison will allow each of the code users to realistically evaluate and improve the organic iodide behaviour sub-models within their codes. (author)

  7. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E., E-mail: luisen.herranz@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Garcia, Monica, E-mail: monica.gmartin@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Morandi, Sonia, E-mail: sonia.morandi@rse-web.it [Nuclear and Industrial Plant Safety Team, Power Generation System Department, RSE, via Rubattino 54, 20134 Milano (Italy)

    2013-12-15

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have

  8. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Morandi, Sonia

    2013-01-01

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have been adopted so that

  9. Water evaporation over sump surface in nuclear containment studies: CFD and LP codes validation on TOSQAN tests

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Degrees du Lou, O. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Arts et Métiers ParisTech, DynFluid Lab. EA92, 151, boulevard de l’Hôpital, 75013 Paris (France); Gelain, T. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France)

    2013-10-15

    Highlights: • Simulations of evaporative TOSQAN sump tests are performed. • These tests are under air–steam gas conditions with addition of He, CO{sub 2} and SF{sub 6}. • ASTEC-CPA LP and TONUS-CFD codes with UDF for sump model are used. • Validation of sump models of both codes show good results. • The code–experiment differences are attributed to turbulent gas mixing modeling. -- Abstract: During the course of a severe accident in a Nuclear Power Plant, water can be collected in the sump containment through steam condensation on walls and spray systems activation. The objective of this paper is to present code validation on evaporative sump tests performed on TOSQAN facility. The ASTEC-CPA code is used as a lumped-parameter code and specific user-defined-functions are developed for the TONUS-CFD code. The seven tests are air–steam tests, as well as tests with other non-condensable gases (He, CO{sub 2} and SF{sub 6}) under steady and transient conditions (two depressurization tests). The results show a good agreement between codes and experiments, indicating a good behavior of the sump models in both codes. The sump model developed as User-Defined Functions (UDF) for TONUS is considered as well validated and is ‘ready-to-use’ for all CFD codes in which such UDF can be added. The remaining discrepancies between codes and experiments are caused by turbulent transport and gas mixing, especially in the presence of non-condensable gases other than air, so that code validation on this important topic for hydrogen safety analysis is still recommended.

  10. Simulation of spray phenomena using the containment code system COCOSYS. 1{sup st} Technical report.Validation and interpretation of selected models and of the coupling of the system codes ATHLET-CD and COCOSYS (VAMKoS); Simulation von Spruehstrahlphaenomenen mit dem Containment Code System COCOSYS. 1. Technischer Fachbericht. Validierung und Analyse ausgewaehlter Modelle sowie der Kopplung der Systemcodes ATHLET-CD und COCOSYS (VAMKoS)

    Energy Technology Data Exchange (ETDEWEB)

    Risken, Tobias; Koch, Marco K.

    2014-12-15

    The present report is the first Technical Report within the research project ''Validation and interpretation of selected models and of the coupling of the system codes ATHLET-CD and COCOSYS'', funded by the German Federal Ministry for Economic Affairs and Energy (BMWi 1501465) and projected at the Reactor Simulation and Safety Group, Chair of Energy Systems and Energy Economics (LEE) at the Ruhr-Universitaet Bochum (RUB). This report deals with the simulation of spray phenomena with the containment code system COCOSYS, which is developed by the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH. First, post-test calculations of the OECD THAI-2 tests HD-30 and HD-31 are presented. The simulation results are compared to experimental values and thereby assessed. The analysis focuses an the assessment of the simultaneous use of the COCOSYS models IVO (spray model) and FRONT (combustion model) as well as a spray entrainment model developed at RUB regarding the simulation of the phenomena related to the interaction of spray and combustion processes. The simulation results show the necessity to consider the induced turbulences. The simulation of these turbulences is performed by modifying the FRONT input parameters leading to an improvement of the simulation results. The consideration of the entrainment positively influences the simulated flow pattern. Subsequently, the simulation of the entrainment of interacting sprays, as occurring in containment spray systems, is considered. The entrainment of interacting sprays is influenced by droplet collisions and changes of the drag between the droplets and the atmosphere. For the simulation an entrainment factor, which has to be determined externally, is implemented into COCOSYS. Exemplary simulations of the OECD SETH-2 ST3 tests show, that in general the use of entrainment factors enables the calculation of alternated gas distributions.

  11. Comparison of computer codes related to the sodium oxide aerosol behavior in a containment building

    International Nuclear Information System (INIS)

    Fermandjian, J.

    1984-09-01

    In order to ensure that the problems of describing the physical behavior of sodium aerosols, during hypothetical fast reactor accidents, were adequately understood, a comparison of the computer codes (ABC/INTG, PNC, Japan; AEROSIM, UKAEA/SRD, United Kingdom; PARDISEKO IIIb, KfK, Germany; AEROSOLS/A2 and AEROSOLS/B1, CEA France) was undertaken in the frame of the CEC: exercise in which code users have run their own codes with a prearranged input

  12. Validation of the containment code Sirius: interpretation of an explosion experiment on a scale model

    International Nuclear Information System (INIS)

    Blanchet, Y.; Obry, P.; Louvet, J.; Deshayes, M.; Phalip, C.

    1979-01-01

    The explicit 2-D axisymmetric Langrangian code SIRIUS, developed at the CEA/DRNR, Cadarache, deals with transient compressive flows in deformable primary tanks with more or less complex internal component geometries. This code has been subjected to a two-year intensive validation program on scale model experiments and a number of improvements have been incorporated. This paper presents a recent calculation of one of these experiments using the SIRIUS code, and the comparison with experimental results shows the encouraging possibilities of this Lagrangian code

  13. [Standards for treatment in forensic committment according to § 63 and § 64 of the German criminal code : Interdisciplinary task force of the DGPPN].

    Science.gov (United States)

    Müller, J L; Saimeh, N; Briken, P; Eucker, S; Hoffmann, K; Koller, M; Wolf, T; Dudeck, M; Hartl, C; Jakovljevic, A-K; Klein, V; Knecht, G; Müller-Isberner, R; Muysers, J; Schiltz, K; Seifert, D; Simon, A; Steinböck, H; Stuckmann, W; Weissbeck, W; Wiesemann, C; Zeidler, R

    2017-08-01

    People who have been convicted of a crime due to a severe mental disorder and continue to be dangerous as a result of this disorder may be placed in a forensic psychiatric facility for improvement and safeguarding according to § 63 and § 64 of the German Criminal Code (StGB). In Germany, approximately 9000 patients are treated in clinics for forensic psychiatry and psychotherapy on the basis of § 63 of the StGB and in withdrawal centers on the basis of § 64 StGB. The laws for treatment of patients in forensic commitment are passed by the individual States, with the result that even the basic conditions differ in the individual States. While minimum requirements have already been published for the preparation of expert opinions on liability and legal prognosis, consensus standards for the treatment in forensic psychiatry have not yet been published. Against this background, in 2014 the German Society for Psychiatry and Psychotherapy, Psychosomatics and Neurology (DGPPN) commissioned an interdisciplinary task force to develop professional standards for treatment in forensic psychiatry. Legal, ethical, structural, therapeutic and prognostic standards for forensic psychiatric treatment should be described according to the current state of science. After 3 years of work the results of the interdisciplinary working group were presented in early 2017 and approved by the board of the DGPPN. The standards for the treatment in the forensic psychiatric commitment aim to initiate a discussion in order to standardize the treatment conditions and to establish evidence-based recommendations.

  14. A research on verification of the CONTAIN CODE model and the uncertainty reduction method for containment integrity

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Moo Hwan; Kang, Seok Hun; Seo, Kyoung Woo [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    1999-03-15

    The final goal of this research is to verify methodology for evaluating more accurately the integrity of containment and develop the methodology to reduce the uncertainty using the data of the operating PWR, KSNPP, KNGR during a severe accident. Therefore, the research selected an indispensable factor about DCH, and analysed sensitivity test at this year.

  15. 9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Canning with heat processing and hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... IDENTIFICATION AS TO CLASS, QUALITY, QUANTITY, AND CONDITION Inspection Procedure § 355.25 Canning with heat...

  16. [Sample German LAPS.

    Science.gov (United States)

    Rosenthal, Bianca

    Four learning activity packages (LAPS) for use in secondary school German programs contain instructional materials which enable students to improve their basic linguistic skills. The units include: (1) "Grusse," (2) "Ich Heisse...Namen," (3) "Tune into Your Career: Business Correspondence 'Auf Deutch'," and (4) "Understanding German Culture."…

  17. Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

    Directory of Open Access Journals (Sweden)

    Omid Noori-Kalkhoran

    2016-10-01

    Full Text Available Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model. In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

  18. Benchmark of the HDR E11.2 containment hydrogen mixing experiment using the MAAP4 code

    International Nuclear Information System (INIS)

    Lee, Sung, Jin; Paik, Chan Y.; Henry, R.E.

    1997-01-01

    The MAAP4 code was benchmarked against the hydrogen mixing experiment in a full-size nuclear reactor containment. This particular experiment, designated as E11.2, simulated a small loss-of-coolant-accident steam blowdown into the containment followed by the release of a hydrogen-helium gas mixture. It also incorporated external spray cooling of the steel dome near the end of the transient. Specifically, the objective of this bench-mark was to demonstrate that MAAP4, using subnodal physics, can predict an observed gas stratification in the containment

  19. Further development of KAVERN and code development on gas generation from the containment basement during concrete decomposition

    International Nuclear Information System (INIS)

    Schwarzott, W.; Artnik, J.; Hassmann, K.; Kemner, H.; Stuckenberg, X.

    1983-04-01

    The events during the melt/concrete interaction, e.g. the shape of the cavity and the mass and energy of gases released to the containment atmosphere can be analysed by the computer code KAVERN. In case of basaltic conrete sump water contacts the melt surface after 7 hours. Overpressurization of the containment is calculated to occur after appr. 5 days. For different paths out of the reactor cavity to the containment atmosphere STROMI calculates the mass flow of the gases released during melt concrete interaction. Results show max. temperatures up to 1200 0 C which is well above the self ignition temperature of H 2 . (orig.) [de

  20. What Information is Stored in DNA: Does it Contain Digital Error Correcting Codes?

    Science.gov (United States)

    Liebovitch, Larry

    1998-03-01

    The longest term correlations in living systems are the information stored in DNA which reflects the evolutionary history of an organism. The 4 bases (A,T,G,C) encode sequences of amino acids as well as locations of binding sites for proteins that regulate DNA. The fidelity of this important information is maintained by ANALOG error check mechanisms. When a single strand of DNA is replicated the complementary base is inserted in the new strand. Sometimes the wrong base is inserted that sticks out disrupting the phosphate backbone. The new base is not yet methylated, so repair enzymes, that slide along the DNA, can tear out the wrong base and replace it with the right one. The bases in DNA form a sequence of 4 different symbols and so the information is encoded in a DIGITAL form. All the digital codes in our society (ISBN book numbers, UPC product codes, bank account numbers, airline ticket numbers) use error checking code, where some digits are functions of other digits to maintain the fidelity of transmitted informaiton. Does DNA also utitlize a DIGITAL error chekcing code to maintain the fidelity of its information and increase the accuracy of replication? That is, are some bases in DNA functions of other bases upstream or downstream? This raises the interesting mathematical problem: How does one determine whether some symbols in a sequence of symbols are a function of other symbols. It also bears on the issue of determining algorithmic complexity: What is the function that generates the shortest algorithm for reproducing the symbol sequence. The error checking codes most used in our technology are linear block codes. We developed an efficient method to test for the presence of such codes in DNA. We coded the 4 bases as (0,1,2,3) and used Gaussian elimination, modified for modulus 4, to test if some bases are linear combinations of other bases. We used this method to analyze the base sequence in the genes from the lac operon and cytochrome C. We did not find

  1. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Beonio-Brocchieri, F.

    1986-09-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes used in reactor safety in order to assess their capability of realistically describing the aerosol behavior in PWR reactor containment buildings during severe accidents. The codes included in the present study are the following: AEROSIM-M, AEROSOLS/Bl, CORRAL-2, NAUA Mod5. In AEROSIM-M, AEROSOLS/Bl and NAUA Mod5, the integro-differential equation for the evolution of the particle mass distribution is approximated by a set of coupled first order differential equations. To this end, the particle distribution function is replaced by a number of discrete monodisperse fractions. The CORRAL-2 has an essentially empirical basis (processes not explicitely modelled, but their net effects accounted for). The physical processes taken into account in the codes are shown finally

  2. Analysis of mitigating measures during steam/hydrogen distributions in nuclear reactor containments with the 3D field code gasflow

    International Nuclear Information System (INIS)

    Royl, P.; Travis, J.R.; Haytcher, E.A.; Wilkening, H.

    1997-01-01

    This paper reports on the recent model additions to the 3D field code GASFLOW and on validation and application analyses for steam/hydrogen transport with inclusion of mitigation measures. The results of the 3D field simulation of the HDR test E11.2 are summarized. Results from scoping analyses that simulate different modes of CO2 inertization for conditions from the HDR test T31.5 are presented. The last part discusses different ways of recombiner modeling during 3D distribution simulations and gives the results from validation calculations for the HDR recombiner test E11.8.1 and the Battelle test MC3. The results demonstrate that field code simulations with computer codes like GASFLOW are feasible today for complex containment geometries and that they are necessary for a reliable prediction of hydrogen/steam distribution and mitigation effects. (author)

  3. New system for the container conditioning of liquid waste in the German future finale repository 'Schacht Konrad'

    International Nuclear Information System (INIS)

    Starke, H.

    2012-01-01

    The full text of publication follows. On-site the NPP Gundremmingen liquid radioactive waste from the NPP water treatment plant is stored in resin or concentrate collecting tanks. These liquid wastes are cemented in containers in order to temporarily store them in the Bavarian interim storage Mitterteich until they are transported into final repository in 'Schacht Konrad'. With this new system liquid radioactive waste is for the first time conditioned directly into containers destined for final repository in 'Schacht Konrad'. Thus, a very secure and sustainable procedure was developed which also provides high profitability. The conditioning plant for resins and concentrate extracts the liquid waste from the respective collecting tank and transports the waste to the separation tank. This separation tank is dimensioned to ensure complete filling of a Konrad container with only one batch. Within the tank there is the option to adjust the suspensions solids content by either extracting supernatant water or by adding de-ionised water. The specific activity is analysed and after the radiologic data and the solids content are available, the containers are cemented. The required amount of cement is based on the solids content and is automatically added. In the mixer, cement and primary waste suspension are mixed. This mixture is filled into the Konrad container via the allocator. The allocator is a funnel-shaped inlet equipped with a movable tube which makes sure the mixture is evenly spread and also ensures optimal filling of the Konrad container. While filling is ongoing, the container is covered by a lowerable splash guard to avoid contamination. The room situation in Gundremmingen and the specific activities of the primary waste suspension make it necessary to disperse the plant to several rooms. Main components such as separation tanks and pumps are installed in shielded rooms. All activities are conducted remotely controlled and are supervised from the central

  4. A mathematical model, and code LIXY, for leaching of radionuclides from containment

    International Nuclear Information System (INIS)

    Fraser, J.L.; Jarvis, R.G.

    1985-06-01

    A mathematical model has been developed to describe the leaching of a radionuclide from an inner region into an outer region, by diffusion processes. Answers have been obtained for the whole range of time values, and have been written into a code LIXY, to calculate the concentration of nuclide at the outer face of the outer region

  5. MISTRA facility for containment lumped parameter and CFD codes validation. Example of the International Standard Problem ISP47

    International Nuclear Information System (INIS)

    Tkatschenko, I.; Studer, E.; Paillere, H.

    2005-01-01

    During a severe accident in a Pressurized Water Reactor (PWR), the formation of a combustible gas mixture in the complex geometry of the reactor depends on the understanding of hydrogen production, the complex 3D thermal-hydraulics flow due to gas/steam injection, natural convection, heat transfer by condensation on walls and effect of mitigation devices. Numerical simulation of such flows may be performed either by Lumped Parameter (LP) or by Computational Fluid Dynamics (CFD) codes. Advantages and drawbacks of LP and CFD codes are well-known. LP codes are mainly developed for full size containment analysis but they need improvements, especially since they are not able to accurately predict the local gas mixing within the containment. CFD codes require a process of validation on well-instrumented experimental data before they can be used with a high degree of confidence. The MISTRA coupled effect test facility has been built at CEA to fulfil this validation objective: with numerous measurement points in the gaseous volume - temperature, gas concentration, velocity and turbulence - and with well controlled boundary conditions. As illustration of both experimental and simulation areas of this topic, a recent example in the use of MISTRA test data is presented for the case of the International Standard Problem ISP47. The proposed experimental work in the MISTRA facility provides essential data to fill the gaps in the modelling/validation of computational tools. (author)

  6. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

    International Nuclear Information System (INIS)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

    2013-10-01

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

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

    International Nuclear Information System (INIS)

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

    1979-07-01

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

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

    International Nuclear Information System (INIS)

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

    1976-11-01

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

  9. Analysis of L test series of ACE (Advanced Containment Experiments) project with modified corcon UW code

    International Nuclear Information System (INIS)

    Laguna Velasco, H.

    1994-01-01

    A series of experimental tests (so call L, Large scale) have been performance under sponsored of many research institutions around the world and management by Electric Power Research Institute at U.S.A. The goal of these tests is to analyze the phenomena of core-concrete interaction at the same conditions as severe accident in light water nuclear reactor. Results of these tests provides experimental data about thermohydraulic phenomenon and aerosol and fission products release. With these results, improves many codes that already have been developed to simulate core-concrete interaction during severe accident ; in case of CORCON.UW code is a improved version developed in University of Wisconsin at CORCON MOD 2. Scope of this work is shown results obtained from CORCON.UW improved. The improves consist of add data about BaSiO 3 , Ba 2 SiO 4 , BaZrO 3 , SrSiO 4 and SrZrO 3 , append Kutateladze's heat transfer correlation, and finally make more efficient the resolution of energy equations system through use a better algorithm. The results obtained by this improved code to the downward power and H 2 , H 2 O, CO and CO 2 release are agree with experimental results, and also it saved 40% of C.P.U. consumption during execution, due improve of energy equation system. Conclusions are, the increase of thermodynamics data in CORCON.UW produce a well results comparative with experimental results and update heat transfer correlations and algorithm brings a versatile code and reliable results. (Author)

  10. Preliminary calculation with code CONTEMPT-LT for spray cooling tests with JAERI model containment vessel

    International Nuclear Information System (INIS)

    Tanaka, Mitsugu

    1978-01-01

    LWR plants have a containment spray system to reduce the escape of radioactive material to the environment in a loss-of-coolant accident (LOCA) by washing out fission products, especially radioiodine, and condensing the steam to lower the pressure. For carrying out the containment spray tests, pressure and temperature behaviour of the JAERI Model Containment Vessel in spray cooling has been calculated with computer program CONTEMPT-LT. The following could be studied quantitatively: (1) pressure and temperature raise rates for steam addition rate and (2) pressure fall rate for spray flow rate and spray heat transfer efficiency. (auth.)

  11. Structural integrity assessment of a pressure container component. Design and service code implementation. Case studies

    International Nuclear Information System (INIS)

    Sanzi, H.C.

    2006-01-01

    In the present work, the most important results of the local stresses occurred in the cracked pipes with a axial through-wall crack (outer), produced during operation of a Petrochemical Plant, using finite elements method, are presented. As requested, the component has been verified based 3D FE plastic analysis, under the postulated failure loading, assuring with this method a high degree of accuracy in the results. Codes used by Design and Service, as ASME Section VIII Div. 2 and API 579, have been used in the analysis. (author) [es

  12. Progress toward NuPack, the ASME code for Type B containments

    International Nuclear Information System (INIS)

    Turula, P.

    1995-01-01

    This paper presented a brief status report on the development of an ASME Code Division for nuclear packaging and discussed some of the more interesting policy decisions as to what is and is not covered in terms of analytical methods, criteria, scope, and other aspects. The process of the development of this Division has been very slow and inconsistent. There were many participants with many diverse interests. The Division 3 rules are close to being ready to be issued. They are a compromise between many needs and the result is certainly not perfect. Opportunities for fine tuning and expanding this document will present themselves after it is issued as future needs become clear

  13. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Castillo G, F.

    2015-01-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  14. Simulation of KAEVER experiments on aerosol behavior in a nuclear power plant containment at accident conditions with the ASTEC code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2006-01-01

    Experiments on aerosol behaviour in saturated and non-saturated atmosphere, which were performed in the KAEVER experimental facility, were simulated with the severe accident computer code ASTEC CPA V1.2. The specific purpose of the work was to assess the capability of the code to model aerosol condensation and deposition in the containment of a light-water-reactor nuclear power plant at severe accident conditions, if the atmosphere saturation conditions are simulated adequately. Five different tests were first simulated with boundary conditions, obtained from the experiments. In all five tests, a non-saturated atmosphere was simulated, although, in four tests, the atmosphere was allegedly saturated. The simulations were repeated with modified boundary conditions, to obtain a saturated atmosphere in all tests. Results of dry and wet aerosol concentrations in the test vessel atmosphere for both sets of simulations are compared to experimental results. (author)

  15. Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

    International Nuclear Information System (INIS)

    Whitesides, G.H.; Stephens, M.E.

    1984-01-01

    During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

  16. Large-scale, multi-compartment tests in PANDA for LWR-containment analysis and code validation

    International Nuclear Information System (INIS)

    Paladino, Domenico; Auban, Olivier; Zboray, Robert

    2006-01-01

    The large-scale thermal-hydraulic PANDA facility has been used for the last years for investigating passive decay heat removal systems and related containment phenomena relevant for next-generation and current light water reactors. As part of the 5. EURATOM framework program project TEMPEST, a series of tests was performed in PANDA to experimentally investigate the distribution of hydrogen inside the containment and its effect on the performance of the Passive Containment Cooling System (PCCS) designed for the Economic Simplified Boiling Water Reactor (ESBWR). In a postulated severe accident, a large amount of hydrogen could be released in the Reactor Pressure Vessel (RPV) as a consequence of the cladding Metal- Water (M-W) reaction and discharged together with steam to the Drywell (DW) compartment. In PANDA tests, hydrogen was simulated by using helium. This paper illustrates the results of a TEMPEST test performed in PANDA and named as Test T1.2. In Test T1.2, the gas stratification (steam-helium) patterns forming in the large-scale multi-compartment PANDA DW, and the effect of non-condensable gas (helium) on the overall behaviour of the PCCS were identified. Gas mixing and stratification in a large-scale multi-compartment system are currently being further investigated in PANDA in the frame of the OECD project SETH. The testing philosophy in this new PANDA program is to produce data for code validation in relation to specific phenomena, such as: gas stratification in the containment, gas transport between containment compartments, wall condensation, etc. These types of phenomena are driven by buoyant high-momentum injections (jets) and/or low momentum injection (plumes), depending on the transient scenario. In this context, the new SETH tests in PANDA are particularly valuable to produce an experimental database for code assessment. This paper also presents an overview of the PANDA SETH tests and the major improvements in instrumentation carried out in the PANDA

  17. International standard problem (ISP) No. 41. Containment iodine computer code exercise based on a radioiodine test facility (RTF) experiment

    International Nuclear Information System (INIS)

    2000-04-01

    International Standard Problem (ISP) exercises are comparative exercises in which predictions of different computer codes for a given physical problem are compared with each other or with the results of a carefully controlled experimental study. The main goal of ISP exercises is to increase confidence in the validity and accuracy of the tools, which were used in assessing the safety of nuclear installations. Moreover, they enable code users to gain experience and demonstrate their competence. The ISP No. 41 exercise, computer code exercise based on a Radioiodine Test Facility (RTF) experiment on iodine behaviour in containment under severe accident conditions, is one of such ISP exercises. The ISP No. 41 exercise was borne at the recommendation at the Fourth Iodine Chemistry Workshop held at PSI, Switzerland in June 1996: 'the performance of an International Standard Problem as the basis of an in-depth comparison of the models as well as contributing to the database for validation of iodine codes'. [Proceedings NEA/CSNI/R(96)6, Summary and Conclusions NEA/CSNI/R(96)7]. COG (CANDU Owners Group), comprising AECL and the Canadian nuclear utilities, offered to make the results of a Radioiodine Test Facility (RTF) test available for such an exercise. The ISP No. 41 exercise was endorsed in turn by the FPC (PWG4's Task Group on Fission Product Phenomena in the Primary Circuit and the Containment), PWG4 (CSNI Principal Working Group on the Confinement of Accidental Radioactive Releases), and the CSNI. The OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) has sponsored forty-five ISP exercises over the last twenty-four years, thirteen of them in the area of severe accidents. The criteria for the selection of the RTF test as a basis for the ISP-41 exercise were; (1) complementary to other RTF tests available through the PHEBUS and ACE programmes, (2) simplicity for ease of modelling and (3) good quality data. A simple RTF experiment performed under controlled

  18. Specific model for a gas distribution analysis in the containment at Almaraz NPP using GOTHIC computer code

    International Nuclear Information System (INIS)

    García González, M.; García Jiménez, P.; Martínez Domínguez, F.

    2016-01-01

    To carry out an analysis of the distribution of gases within the containment building at the CN Almaraz site, a simulation model with the thermohydraulic GOTHIC [1] code has been used. This has been assessed with a gas control system based on passive autocatalytic recombiners (PARs). The model is used to test the effectiveness of the control systems for gases to be used in the Almaraz Nuclear Power Plant, Uits I&II (Caceres, Spain, 1,035 MW and 1,044 MW). The model must confirm the location and number of the recombiners proposed to be installed. It is an essential function of the gas control system to avoid any formation of explosive atmospheres by reducing and limiting the concentration of combustible gases during an accident, thus maintaining the integrity of the containment. The model considers severe accident scenarios with specific conditions that produce the most onerous generation of combustible gases.

  19. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  20. Evaluation of passive autocatalytic recombiners (PARS) performance for a PWR-konvoi containment type with Gothic 8.1 code

    International Nuclear Information System (INIS)

    Lopez-Alonso Conty, E.; Papini, D.; Jimenez Varas, G.

    2016-01-01

    The study presented in this work analyses the evaluation of Passive Autocatalytic Recombiners (PARs) performance for a PWR-Konvoi containment type as a result of an international collaboration between the Paul Scherrer institute (PSI) and the Universidad Politecnica de Madrid (UPM). The implementation study analyzes the size, location and number of the PARs to minimize the risk arising from a hydrogen release and its distribution in the containment building during a hypothetical severe accident. A detailed 3D model of containment was used for the simulations developed for the Gothic 8.1 code. In the first place, the hydrogen preferential pathways and points of hydrogen accumulation were studies and identified starting from the base case scenario without any mitigation measure. The severe accident scenario chosen is a fast release of hydrogen-steam mixture from hot leg creep rupture during SBO (Station Black-Out) accident. Secondly a configuration of PARs was simulated under the same conditions of the unmitigated case. The PAR configuration offered an improvement in the chosen accident scenario, decreasing the hydrogen concentration values below the flammability limit /hydrogen concentration below 7%) in all the containment compartments. (Author)

  1. Containers analysis code of zero order (CACO0) - A basic design system for Type B packages

    International Nuclear Information System (INIS)

    Gaspar, C.; Benito, G.; Rey, J.C.

    1989-01-01

    Very frequently, the principal issues that have to be assessed in the design of a type B(U) package are radiation shielding and evaluation of mechanical and thermal test effects. Thermal behavior during normal transport conditions has also to be considered when the material must dissipate high thermal power. If the transported material is fissile it should be assured that it remains subcritical during transport. The containment of radioactive material must always be assured. In some cases this requires considerable effort. Usually these different design issues are very closely coupled. This coupling does not permit independent consideration. Also, some issues are competitive and generate conflicting design criteria. Given the goal of meeting pertinent transport regulations at a reasonable cost, all design-relevant issues must be balanced in order to obtain a good design. For each design-relevant issue there exists a number of methods of varying efficiency and cost, which can be used to define the key parameters of those particular issues. The overall design methodology must taken into account interactions between parameters of different issues. CACO0 is a system that integrates all design relevant issues and their interactions. The system consists of different modules, each one oriented to a different design issue. The modules are related by a control structure that enables sequentation or iteration during design in a fast and simple manner. Modules can easily be replaced or added, so the system can be updated or adapted to new design problems. The system was designed for use in factibility analysis, cost estimation, conceptual design and initial stages of basic design of type B(U) packages. To accomplish those ends, simple, fast and conservative methods are used

  2. Assessment of Ultimate Load Capacity for Pre-Stressed Concrete Containment Vessel Model of PWR Design With BARC Code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Singh, R.K.; Patnaik, R.; Ramanujam, S.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian Pressurised Heavy Water Reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results. The present paper highlights the analysis results for Prestressed Concrete Containment Vessel (PCCV) tested at Sandia National Labs, USA in a Round Robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd= design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete-tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd. (authors)

  3. GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

    International Nuclear Information System (INIS)

    Ramachandran, S.; Krause, M.; Nguyen, T.

    2003-01-01

    This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

  4. Computer simulations of a generic truck cask in a regulatory fire using the Container Analysis Fire Environment (CAFE) code

    International Nuclear Information System (INIS)

    Ju, H.; Greiner, M.; Suo-Anttila, A.

    2002-01-01

    The Container Analysis Fire Environment (CAFE) computer code is designed to predict accurately convection and radiation heat transfer to a thermally massive object engulfed in a large pool fire. It is well suited for design and risk analyses of spent nuclear fuel transport systems. CAFE employs computational fluid dynamics and several fire and radiation models. These models must be benchmarked using experimental results. In this paper, a set of wind velocity conditions are determined which allow CAFE accurately to reproduce recent heat transfer measurements for a thick walled calorimeter in a ST-1 regulatory pool fire. CAFE is then used to predict the response of an intack (thin walled) generic legal weight truck cask. The maximum temperatures reached by internal components are within safe limits. A simple 800 deg. C, grey-radiation fire model gives maximum component temperatures that are somewhat below those predicted by CAFE. (author)

  5. Development of the Computer Code to Determine an Individual Radionuclides in the Rad-wastes Container for Ulchin Units 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Kang, D.W.; Chi, J.H.; Goh, E.O. [Korea Electric Power Research Institute, Taejon (Korea)

    2001-07-01

    A computer program, RASSAY was developed to evaluate accurately the activities of various nuclides in the rad-waste container for Ulchin units 3 and 4. This is the final report of the project, {sup D}evelopment of the Computer Code to Determine an Individual Radionuclides in the Rad-wastes Container for Ulchin Units 3 and 4 and includes the followings; 1) Structure of the computer code, RASSAY 2) An example of surface dose calculation by computer simulation using MCNP code 3) Methods of sampling and activity measurement of various Rad-wastes. (author). 21 refs., 35 figs., 6 tabs.

  6. German atomic low meeting 2004

    International Nuclear Information System (INIS)

    Ossenbuehl, F.

    2005-01-01

    The conference report on the German atomic law meeting 2004 contains 14 contributions on the German atomic legislation within four parts: Damage precaution in the operational phase; Legal general requirements for the final disposal - considerations ''de lege lata'' and ''de lege ferenda''. Financing of the site searching by a statutory company (''Verbandsmodell''). Atomic supervision authority - federal executive administration or federal self administration?

  7. MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae

    1998-01-01

    The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others

  8. MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae [KAERI, Taejon (Korea, Republic of)

    1998-10-01

    The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others.

  9. Environmental protection and penal law in Greece - a comparison with the German penal code on environmental matters. Der strafrechtliche Umweltschutz in Griechenland unter besonderer Beruecksichtigung des Deutschen Umweltstrafrechts

    Energy Technology Data Exchange (ETDEWEB)

    Karamanidis, G.

    1985-01-01

    The first chapter outlines the ecological situation of Greece, while the second chapter presents the legal foundations of environmental protection in Greece. Secondary laws are mentioned, as these are generally the laws in which penal liabilities are stated. The present environmental protection regulations are found to be unsatisfactory and unfit for preventing environmental damage. A new legislative structure is proposed on the basis of German environmental protection standards. (orig./HSCH).

  10. German Vocabulary.

    Science.gov (United States)

    Coombs, Virginia M.

    This article discusses in general terms derivational aspects of English vocabulary. Citing examples of Anglo-Saxon origin, the author provides a glimpse into the nature of the interrelatedness of English, German, and French vocabulary. (RL)

  11. German Orientalism

    OpenAIRE

    Margaret Olin

    2011-01-01

    Review of: Suzanne L. Marchand, German Orientalism in the Age of Empire: Religion, Race and Scholarship, Cambridge and Washington, D.C.: Cambridge University Press, 2009. This analysis of Suzanne L. Marchand’s German Orientalism in the Age of Empire: Religion, Race and Scholarship reads her contribution in part against the background of Edward Said’s path breaking book Orientalism. Differences lie in her more expansive understanding of the term ‘Oriental’ to include the Far East and her conce...

  12. Layers of root nouns in Germanic

    DEFF Research Database (Denmark)

    Hansen, Bjarne Simmelkjær Sandgaard

    2017-01-01

    The root-noun declension became productive in early Germanic, containing (I) inherited root nouns, (IIa) original substrate or loan words, and transitions from other declensions in (IIb) Proto-Germanic and (III) North Germanic. As ablaut was abolished, the inherited type would display ablaut grades...

  13. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

    International Nuclear Information System (INIS)

    Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

    1987-11-01

    An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

  14. ZOCO VI - a computer code to calculate the time- and space-dependent pressure distribution in full pressure containments of water-cooled reactors

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1974-12-01

    ZOCO VI is a computer code to investigate the time and space dependent pressure distribution in full pressure containment of water cooled nuclear power reactors following a loss-of-coolant accident, which is caused by the rupture of a main coolant or steam line. ZOCO VI is an improved version of the computer code ZOCO V with enlarged description of condensing events. (orig.) [de

  15. BAR-CODE BASED WEIGHT MEASUREMENT STATION FOR PHYSICAL INVENTORY TAKING OF PLUTONIUM OXIDE CONTAINERS AT THE MINING AND CHEMICAL COMBINE RADIOCHEMICAL REPROCESSING PLANT NEAR KRASNOYARSK, SIBERIA

    International Nuclear Information System (INIS)

    SUDA, S.

    1999-01-01

    This paper describes the technical tasks being implemented to computerize the physical inventory taking (PIT) at the Mining and Chemical Combine (Gorno-Khimichesky Kombinat, GKhK) radiochemical plant under the US/Russian cooperative nuclear material protection, control, and accounting (MPC and A) program. Under the MPC and A program, Lab-to-Lab task agreements with GKhK were negotiated that involved computerized equipment for item verification and confirmatory measurement of the Pu containers. Tasks under Phase I cover the work for demonstrating the plan and procedures for carrying out the comparison of the Pu container identification on the container with the computerized inventory records. In addition to the records validation, the verification procedures include the application of bar codes and bar coded TIDs to the Pu containers. Phase II involves the verification of the Pu content. A plan and procedures are being written for carrying out confirmatory measurements on the Pu containers

  16. Consideration of turbulent deposition in aerosol behaviour modelling with the CONTAIN code and comparison of the computations to sodium release experiments

    International Nuclear Information System (INIS)

    Jonas, R.

    1988-09-01

    CONTAIN is a computer code to analyze physical, chemical and radiological processes inside the reactor containment in the sequence of severe reactor accident. Modelling of the aerosol behaviour is included. We have improved the code by implementing a subroutine for turbulent deposition of aerosols. In contrast to previous calculations in which this effect was neglected, the computer results are in good agreement with sodium release experiments. If a typical friction velocity of 1 m/s is chosen, the computed aerosol mass median diameters and aerosol mass concentrations agree with the experimental results within a factor of 1.5 or 2, respectively. We have also found a good agreement between the CONTAIN calculations and results from other aerosol codes. (orig.) [de

  17. An analysis, using the CLAPTRAP code, of the pressure transients developed in the Carolinas Virginia Tube Reactor during containment performance tests

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1982-11-01

    To check containment performance of the CVTR, steam was injected above the operating floor through a 10 foot pipe cap containing the 1 inch diameter holes, at a steady rate of 102.8 lb/sec for a period of 166 seconds. This steam had an enthalpy of 1195 Btu/lb and was therefore not entirely typical of the much wetter material which would be rejected for the greater part of a true breached circuit accident. Pressure transients measured experimentally within the containment were compared with results calculated by the American code CONTEMPT and these results in turn have allowed the Winfrith code CLAPTRAP to be tested for consistency and to establish that the use of this code would have led to similar conclusions about the heat transfer coefficients at the heat absorbent surfaces. (U.K.)

  18. 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

  19. Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems, including containment. Summary report of a technical meeting

    International Nuclear Information System (INIS)

    2003-11-01

    Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The publication constitutes the report of the Technical Meeting. It includes short summaries of the presentations that were made and of the discussions as well as conclusions and

  20. 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

    Containment of the WWER-440 Model 213 nuclear power plant features a Bubble Condenser, a complex passive pressure suppression system, intended to limit pressure rise in the containment during accidents. Due to lack of experimental evidence of its successful operation in the original design documentation, the performance of this system under accidents with ruptures of large high-energy pipes of the primary and secondary sides remains a known safety concern for this containment type. Therefore, a number of research and analytical studies have been conducted by the countries operating WWER-440 reactors and their Western partners in the recent years to verify Bubble Condenser operation under accident conditions. Comprehensive experimental research studies at the Electrogorsk BC V-213 test facility, commissioned in 1999 in Electrogorsk Research and Engineering Centre (EREC), constitute essential part of these efforts. Nowadays this is the only operating large-scale facility enabling integral tests on investigation of the Bubble Condenser performance. Several large international research projects, conducted at this facility in 1999-2003, have covered a spectrum of pipe break accidents. These experiments have substantially improved understanding of the overall system performance and thermal hydraulic phenomena in the Bubble Condenser Containment, and provided valuable information for validating containment codes against experimental results. One of the recent experiments, denoted as SLB-G02, has simulated steam line break. The results of this experiment are of especial value for the engineers working in the area of computer code application for WWER-440 containment analyses, giving an opportunity to verify validity of the code predictions and identify possibilities for model improvement. This paper describes the results of the post-test calculations of the SLB-G02 experiment, conducted as a joint effort of GRS, Germany and Ukrainian technical support organizations for

  1. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  2. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  3. Development of 3D models of buildings for containment of the nuclear power plant of Almaraz and of the Trillo Nuclear with the GOTHIC 8.0 code

    International Nuclear Information System (INIS)

    Jimenez, G.; Bocanegra Melian, R.; Fernandez Cosils, K.; Barreira Pereira, P.; Rey Peinado, L.; Posada Barral, J. M.

    2014-01-01

    The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)

  4. Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2016-05-15

    Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

  5. A research on the verification of models used in the computational codes and the uncertainty reduction method for the containment integrity evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Moo Hwan; Seo, Kyoung Woo [POSTECH, Pohang (Korea, Republic of)

    2001-03-15

    In the probability approach, the calculated CCFPs of all the scenarios were zero, which meant that it was expected that for all the accident scenarios the maximum pressure load induced by DCH was lower than the containment failure pressure obtained from the fragility curve. Thus, it can be stated that the KSNP containment is robust to the DCH threat. And uncertainty of computer codes used to be two (deterministic and probabilistic) approaches were reduced by the sensitivity tests and the research with the verification and comparison of the DCH models in each code. So, this research was to evaluate synthetic result of DCH issue and expose accurate methodology to assess containment integrity about operating PWR in Korea.

  6. 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.

  7. [Quality assurance in coding expertise of hospital cases in the German DRG system. Evaluation of inter-rater reliability in MDK expertise].

    Science.gov (United States)

    Huber, H; Brambrink, M; Funk, R; Rieger, M

    2012-10-01

    The purpose of this study was to evaluate differences in the D-DRG results of a hospital case by 2 independently coding MKD raters. Calculation of the 2-inter-rater reliability was performed by examination of the coding of individual hospital cases. The reasons for the non-agreement of the expert evaluations and suggestions to improve the process are discussed. From the expert evaluation pool of the MDK-WL a random sample of 0.7% of the 57,375 expertises was taken. Distribution equality with the basic total was tested by the χ² test or, respectively, Fisher's exact test. For the total of 402 individual hospital cases, the G-DRG case sums of 2 experts of the MDK were determined independently and the results checked for each individual case for agreement or non-agreement. The corresponding confidence intervals with standard errors were analysed to test if certain major diagnosis categories (MDC) were statistically significantly more affected by differing expertise results than others. In 280 of the total 402 tested hospital cases, the 2 MDK raters independently reached the same G-DRG results; in 122 cases the G-DRG case sums determined by the 2 raters differed (agreement 70%; CI 65.2-74.1). Different DRG results between the 2 experts occurred regularly in the entire MDC spectrum. No MDC chapter in which significant differences between the 2 raters arose could be identified. The results of our study demonstrate an almost 70% agreement in the evaluation of hospital cost accounts by 2 independently operating MDK. This result leaves room for improvement. Optimisation potentials can be recognised on the basis of the results. Potential for improvement was established in combination with regular further training and the expansion of binding internal code recommendations as well as exchange of code-relevant information among experts in internal forums. The presented model is in principle suitable for cross-border examinations within the MDK system with the advantage that

  8. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-09-01

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

  9. Simulation of gas mixing and transport in a multi-compartment geometry using the GOTHIC containment code and relatively coarse meshes

    International Nuclear Information System (INIS)

    Andreani, Michele; Paladino, Domenico

    2010-01-01

    The recently concluded OECD SETH project included twenty-four experiments on basic flows and gas transport and mixing driven by jets and plumes in two, large, connected vessels of the PANDA facility. The experiments featured injection of saturated or superheated steam, or a mixture of steam and helium in one vessel and venting from the same vessel or from the connected one. These tests have been especially designed for providing an extensive data base for the assessment of three-dimensional codes, including CFD codes. In particular, one of the goals of the analytical activities associated with the experiments was to evaluate the detail of the model (mesh) necessary for capturing the various phenomena. This work reports an overview of the results obtained for these experimental data using the advanced containment code GOTHIC and relatively coarse meshes, which are coarser than the ones typically used for the simulation with commercial CFD codes, but are still representative of the models which are currently affordable for a full containment analysis. In general, the phenomena were correctly represented in the simulations with GOTHIC, and the agreement of the results with the data was in most cases pretty good, in some cases excellent. Only for a few tests (or particular phenomena occurring in some tests) the simulations showed noticeable discrepancies with the experimental data, which could be referred to either an insufficiently detailed mesh or to lack of specialized models for local effects.

  10. Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

    International Nuclear Information System (INIS)

    Langhans, J.

    1991-01-01

    Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs. (orig.)

  11. Model of the containment building of Almaraz NPP and the system of recombiners PARs, with the GOTHIC code, for the study of the diffusion of combustible gases

    International Nuclear Information System (INIS)

    Garcia Gonzalez, M.; Huelamo, E.; Mazrtinez, M.; Perez, J. R.

    2014-01-01

    This paper presents the analysis of distribution of gases within the containment building carried out a simulation model with the code Thermo hydraulic GOTHIC, which has been evaluated based on passive autocatalytic recombiners gas control system. The model considers scenarios of severe accident with specific conditions that produce the most hydrogen generation rates. Intended to verify the effectiveness of the control system of gas expected to be installed in the Almaraz Nuclear power plant so that the number and location of recombiners equipment meets its function of preventing the formation of explosive atmospheres which impairs the integrity of the containment, reducing and limiting the concentration of combustible gases during the postulated accident. (Author)

  12. Teaching German-Americana

    Science.gov (United States)

    Tolzmann, Don Heinrich

    1976-01-01

    A university course entitled "The German-Americans" attempted to study and evaluate German culture in the U. S. Lecture topics and term paper theses are listed and a selected annotated bibliography of German-American culture is included. (CHK)

  13. 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.

  14. Simulation of buoyancy induced gas mixing tests performed in a large scale containment facility using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Z.; Chin, Y.S. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    This paper compares containment thermal-hydraulics simulations performed using GOTHIC against a past test set of large scale buoyancy induced helium-air-steam mixing experiments that had been performed at the AECL's Chalk River Laboratories. A number of typical post-accident containment phenomena, including thermal/gas stratification, natural convection, cool air entrainment, steam condensation on concrete walls and active local air cooler, were covered. The results provide useful insights into hydrogen gas mixing behaviour following a loss-of-coolant accident and demonstrate GOTHIC's capability in simulating these phenomena. (author)

  15. Simulation of buoyancy induced gas mixing tests performed in a large scale containment facility using GOTHIC code

    International Nuclear Information System (INIS)

    Liang, Z.; Chin, Y.S.

    2014-01-01

    This paper compares containment thermal-hydraulics simulations performed using GOTHIC against a past test set of large scale buoyancy induced helium-air-steam mixing experiments that had been performed at the AECL's Chalk River Laboratories. A number of typical post-accident containment phenomena, including thermal/gas stratification, natural convection, cool air entrainment, steam condensation on concrete walls and active local air cooler, were covered. The results provide useful insights into hydrogen gas mixing behaviour following a loss-of-coolant accident and demonstrate GOTHIC's capability in simulating these phenomena. (author)

  16. Evaluation and Selection of a Multi-Dimensional Code for H{sub 2} Combustion and Explosion Analysis in the Containment of a Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyung Seok; Kim, Sangbaik; Hong, Seongwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Passive Auto-Catalytic Recombiners (PARs) were installed in all NPP containments to reduce hydrogen concentration during a severe accident. However, hydrogen combustion is possible during a severe accident if the hydrogen concentration is higher than about 10% at a local position in the containment. Thus, to assure containment integrity, it is necessary to evaluate an overpressure buildup resulting from a propagation of hydrogen flame along the obstacle and wall in the containment during a severe accident. Korea Atomic Energy Research Institute (KAERI) decided to import the computational code for the hydrogen combustion and explosion analysis from a foreign country, to establish a numerical analysis system for considering hydrogen generation in the core, to hydrogen combustion in the containment, as soon as possible. KAERI chose the COM3D as the computational code for hydrogen combustion and explosion analysis by evaluating for its numerical methods, physical models, a solver algorithm, validation and application results, and its ability to connect GASFLOW for calculating hydrogen distribution. In addition, the COM3D is currently used to evaluate the integrity of the EPR containment by predicting the overpressure buildup resulting from the hydrogen flame acceleration with the validated analysis methodology. However, we have to find a way to transfer the GASFLOW results, with a cylindrical grid model, as the initial condition of COM3D with a Cartesian grid model, because the COM3D can automatically import the GASFLOW result only when the Cartesian grid model is used, whereas KAERI has performed the GASFLOW analysis with the cylindrical grid model.

  17. New German abortion law agreed.

    Science.gov (United States)

    Karcher, H L

    1995-07-15

    The German Bundestag has passed a compromise abortion law that makes an abortion performed within the first three months of pregnancy an unlawful but unpunishable act if the woman has sought independent counseling first. Article 218 of the German penal code, which was established in 1871 under Otto von Bismarck, had allowed abortions for certain medical or ethical reasons. After the end of the first world war, the Social Democrats tried to legalize all abortions performed in the first three months of pregnancy, but failed. In 1974, abortion on demand during the first 12 weeks was declared legal and unpunishable under the social liberal coalition government of chancellor Willy Brandt; however, the same year, the German Federal Constitution Court in Karlsruhe ruled the bill was incompatible with article 2 of the constitution, which guarantees the right to life and freedom from bodily harm to everyone, including the unborn. The highest German court also ruled that a pregnant woman had to seek a second opinion from an independent doctor before undergoing an abortion. A new, extended article 218, which included a clause giving social indications, was passed by the Bundestag. When Germany was unified, East Germans agreed to be governed by all West German laws, except article 218. The Bundestag was given 2 years to revise the article; however, in 1993, the Federal Constitution Court rejected a version legalizing abortion in the first 3 months of the pregnancy if the woman sought counsel from an independent physician, and suggested the recent compromise passed by the Bundestag, the lower house of the German parliament. The upper house, the Bundesrat, where the Social Democrats are in the majority, still has to pass it. Under the bill passed by the Bundestag, national health insurance will pay for an abortion if the monthly income of the woman seeking the abortion falls under a certain limit.

  18. Aminotryptophan-containing barstar: structure--function tradeoff in protein design and engineering with an expanded genetic code.

    Science.gov (United States)

    Rubini, Marina; Lepthien, Sandra; Golbik, Ralph; Budisa, Nediljko

    2006-07-01

    The indole ring of the canonical amino acid tryptophan (Trp) possesses distinguished features, such as sterical bulk, hydrophobicity and the nitrogen atom which is capable of acting as a hydrogen bond donor. The introduction of an amino group into the indole moiety of Trp yields the structural analogs 4-aminotryptophan ((4-NH(2))Trp) and 5-aminotryptophan ((5-NH(2))Trp). Their hydrophobicity and spectral properties are substantially different when compared to those of Trp. They resemble the purine bases of DNA and share their capacity for pH-sensitive intramolecular charge transfer. The Trp --> aminotryptophan substitution in proteins during ribosomal translation is expected to result in related protein variants that acquire these features. These expectations have been fulfilled by incorporating (4-NH(2))Trp and (5-NH(2))Trp into barstar, an intracellular inhibitor of the ribonuclease barnase from Bacillus amyloliquefaciens. The crystal structure of (4-NH(2))Trp-barstar is similar to that of the parent protein, whereas its spectral and thermodynamic behavior is found to be remarkably different. The T(m) value of (4-NH(2))Trp- and (5-NH(2))Trp-barstar is lowered by about 20 degrees Celsius, and they exhibit a strongly reduced unfolding cooperativity and substantial loss of free energy in folding. Furthermore, folding kinetic study of (4-NH(2))Trp-barstar revealed that the denatured state is even preferred over native one. The combination of structural and thermodynamic analyses clearly shows how structures of substituted barstar display a typical structure-function tradeoff: the acquirement of unique pH-sensitive charge transfer as a novel function is achieved at the expense of protein stability. These findings provide a new insight into the evolution of the amino acid repertoire of the universal genetic code and highlight possible problems regarding protein engineering and design by using an expanded genetic code.

  19. Baltic, Slavic, Germanic

    Directory of Open Access Journals (Sweden)

    Frederik Kortlandt

    2017-02-01

    Full Text Available The western Indo-European vocabulary in Baltic and Slavic is the result of an Indo-European substratum which contained an older non-Indo-European layer and was part of the Corded Ware horizon. The numbers show that a considerable part of the vocabulary was borrowed after the split between Baltic and Slavic, which came about when their speakers moved westwards north and south of the Pripet marshes. Germanic and Balto-Slavic were never contiguous Indo-European dialects at any stage of their prehistory.

  20. Code-experiment comparison on wall condensation tests in the presence of non-condensable gases-Numerical calculations for containment studies

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France); Porcheron, E.; Dumay, F.; Vendel, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Steam condensation on walls has been investigated in the TOSQAN vessel. Black-Right-Pointing-Pointer Experiments on 7 different tests have been performed. Black-Right-Pointing-Pointer Different steam injections and wall temperatures are used. Black-Right-Pointing-Pointer Simulations are performed in 2D using the TONUS code. Black-Right-Pointing-Pointer Code-experiments comparisons at many different locations show a good agreement. - Abstract: During the course of a severe Pressurized Water Reactor accident, pressurization of the containment occurs and hydrogen can be produced by the reactor core oxidation and distributed in the containment according to convection flows and wall condensation. Filmwise wall condensation in the presence of non-condensable gases is a subject of many interests and extensive studies have been performed in the past. Some empirical correlations have demonstrated their limit for extrapolation under different thermal-hydraulic conditions and at different geometries/scales. The French Institute for Radiological Protection and Nuclear Safety (IRSN) has developed a numerical tool and an experimental facility in order to investigate free convection flows in the presence of condensation. The objective of this paper is to present numerical results obtained on different wall condensation tests in 7 m{sup 3} volume vessel (TOSQAN facility), and to compare them with the experimental ones. Over eight tests are considered here, and code-experiment comparison is performed on many different locations, giving an extensive insight of the code assessment for air-steam mixture flows involving wall condensation in the presence of non-condensable gases.

  1. Modelling of local steam condensation on walls in presence of noncondensable gases. Application to a local calculation in reactor containment using the multidimensional geyser code

    International Nuclear Information System (INIS)

    Benet, L.V.; Caroli, C.; Cornet, P.; Coulon, N.; Magnaud, J.P.

    1995-01-01

    The frame of this paper is the study of severe accidents of pressurized water reactors (PWR). The need for containment modelling, and in particular for hydrogen risk study, was reinforced in France after 1990, with the requirement of taking into account severe accidents in design of future plants. This new need of assessing the transient local hydrogen concentration incited us to develop the multidimensional code GEYSER for containment analysis. This code, developed by the Department of Mechanics and Technology of the French Atomic Energy Commission, is presented here with a detailed example of calculation. We describe the mixture, whose constituents are noncondensable gases (air or air and hydrogen) and water vapor liquid, by a compressible homogeneous diphasic model. The wall condensation is based on the Chilton-Colburn formulation and heat mass transfer analogy. We present a transient two-dimensional axisymmetric calculation of a simplified accidental sequence during one hour. The calculation in the large volume of a reactor containment consists of an injection of vapor, first important then moderate, followed by an injection of hydrogen. (authors). 8 refs., 4 figs., 4 tabs

  2. Elaboration of data and documents intended to complement and expand the German series of nuclear engineering codes. 4. Technical report. Current knowledge of the mechanisms of corrosion fatigue in ferritic materials

    International Nuclear Information System (INIS)

    Frank, J.

    1997-01-01

    Concerning the processes of crack initiation under cyclic loading in a high-temperature water environment, examinations are performed in the USA for amending the design-basis curves published in ASME section III, which are comparable to those contained in the KTA code 3201.2. A revision of ASME section III has not yet been published. The same applies to the processes of crack propagation under cyclic loading in a high-temperature water environment, US experts examining for the purpose of amendment the limiting curves of cyclic crack propagation rates of carbon and low-alloyed steels in a water environment published in ASME section XI. (Orig./CB) [de

  3. Legal relevance of the purpose of contract in German law

    Directory of Open Access Journals (Sweden)

    Dudaš Atila

    2013-01-01

    Full Text Available Unlike the French Civil code, the German Civil code belongs to the group of so-called anti-causalistic codifications, since it explicitly does not govern the issue of purpose (cause of contract. Due to this very reason, the delineation between abstract and causal juridical acts gains special importance in German law. The German Civil Code governs a number of juridical acts and other acts of legal importance that are abstract in their nature. Among them the abstract nature of the promise to fulfill an obligation (Schuldversprechung and the acknowledgement of a debt (Schuldannerkennung is traditionally considered the most prominent. However, the relation to the purpose for which they are concluded is not entirely interrupted, since in the case of frustration of their purpose, any asset given to the other party is subject to restitution under the rules of unjustified enrichment. The fact that the issue of purpose of contract is not explicitly governed in the German Civil Code, does not lead to the conclusion, though, that it is legally irrelevant. It gains legal relevance in two different aspects: as a licit and as an illicit purpose. On the one hand, juridical acts concluded with the aim to achieve illicit purposes are considered void, for which the Code's sections on the general confines of the principle of freedom of contract serve the statutory basis - such juridical acts infringe the institution of 'good customs' (gute Sitten, usually referred to as public policy, while the performance of other factual or legal acts in order to achieve illicit purposes are sanctioned under the rules of unjustified enrichment. On the other hand, lawful purposes of the parties gain legal relevance in relation to a range of various institutions. Concerning some of them the Code itself contains formulations implying the necessity to ascertain the purpose of contract, while in other cases the case law and the doctrine have come to such conclusion. The determination

  4. Aging plant life management - the requirements defined to date by the KTA nuclear engineering codes

    International Nuclear Information System (INIS)

    Kalinowski, I.

    1996-01-01

    German nuclear engineering codes so far do not enclose a specific aging plant life management programme. However, the existing codes and standards do contain a number of applicable requirements and principles of relevance to objectives and principles of such programmes, as they also cover aging-induced effects on power plants. The major principles relating to preventive safety engineering and quality assurance are laid down in the publications KTA 1401, 1404, 1201, 1202, and KTA 3211. (DG) [de

  5. Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

    Directory of Open Access Journals (Sweden)

    Bruno Gonfiotti

    2017-01-01

    Full Text Available The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0 employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

  6. 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)

  7. Stand-alone containment analysis of Phébus FPT tests with ASTEC and MELCOR codes: the FPT-2 test.

    Science.gov (United States)

    Gonfiotti, Bruno; Paci, Sandro

    2018-03-01

    During the last 40 years, many studies have been carried out to investigate the different phenomena occurring during a Severe Accident (SA) in a Nuclear Power Plant (NPP). Such efforts have been supported by the execution of different experimental campaigns, and the integral Phébus FP tests were probably some of the most important experiments in this field. In these tests, the degradation of a Pressurized Water Reactor (PWR) fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the findings on these and previous tests, numerical codes such as ASTEC and MELCOR have been developed to analyze the evolution of a SA in real NPPs. After the termination of the Phébus FP campaign, these two codes have been furthermore improved to implement the more recent findings coming from different experimental campaigns. Therefore, continuous verification and validation is still necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The aim of the present work is to re-analyze the Phébus FPT-2 test employing the updated ASTEC and MELCOR code versions. The analysis focuses on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel (CV) have been developed. The paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products (FP) behavior. When possible, a comparison among the results obtained during this work and by different authors in previous work is also performed. This paper is part of a series of publications covering the four Phébus FP tests using a PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3, excluding the FPT-4 one, related to the study of the release of low-volatility FP and transuranic elements from a debris bed and a pool of melted fuel.

  8. Stand-alone containment analysis of Phébus FPT tests with ASTEC and MELCOR codes: the FPT-2 test

    Directory of Open Access Journals (Sweden)

    Bruno Gonfiotti

    2018-03-01

    Full Text Available During the last 40 years, many studies have been carried out to investigate the different phenomena occurring during a Severe Accident (SA in a Nuclear Power Plant (NPP. Such efforts have been supported by the execution of different experimental campaigns, and the integral Phébus FP tests were probably some of the most important experiments in this field. In these tests, the degradation of a Pressurized Water Reactor (PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the findings on these and previous tests, numerical codes such as ASTEC and MELCOR have been developed to analyze the evolution of a SA in real NPPs. After the termination of the Phébus FP campaign, these two codes have been furthermore improved to implement the more recent findings coming from different experimental campaigns. Therefore, continuous verification and validation is still necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The aim of the present work is to re-analyze the Phébus FPT-2 test employing the updated ASTEC and MELCOR code versions. The analysis focuses on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel (CV have been developed. The paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products (FP behavior. When possible, a comparison among the results obtained during this work and by different authors in previous work is also performed. This paper is part of a series of publications covering the four Phébus FP tests using a PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3, excluding the FPT-4 one, related to the study of the release of low-volatility FP and transuranic elements from a debris bed and a pool of melted fuel. Keywords: Safety

  9. Simulation of the heat transfer of a irradiated fuel storage container with code CFD STAR- CCM+; Simulacion de la transferencia de calor de un contenedor de almacenamiento de combustible irradiado con el codigo CFD STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Barrera matalla, J. E.; Hernandez Gomez, J.; Riverala Gurruchaga, J.

    2012-07-01

    Irradiated fuel has become an object of interest in the industry by the importance of ensuring its safety during long periods of storage time. New containers, stores, methods and codes will be used to ensure a suitable cooling and residual heat removal, and secure the safety of fuel elements in dry storage. The codes CFD (Computational Fluid Dynamics) have great potential to help in design of containers and stores, improving thermal-hydraulic performance and the extraction of heat generated.

  10. Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

    International Nuclear Information System (INIS)

    Kolev, N.I.; Sybotinov, L.S.

    1984-01-01

    The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

  11. 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.)

  12. Rare earth germanates

    International Nuclear Information System (INIS)

    Bondar', I.A.; Vinogradova, N.V.; Dem'yanets, L.N.

    1983-01-01

    Rare earth germanates attract close attention both as an independent class of compounds and analogues of a widely spread class of natural and synthetic minerals. The methods of rare earth germanate synthesis (solid-phase, hydrothermal) are considered. Systems on the basis of germanium and rare earth oxides, phase diagrams, phase transformations are studied. Using different chemical analysese the processes of rare earth germanate formation are investigated. IR spectra of alkali and rare earth metal germanates are presented, their comparative analysis being carried out. Crystal structures of the compounds, lattice parameters are studied. Fields of possible application of rare earth germanates are shown

  13. Rare earth germanates

    International Nuclear Information System (INIS)

    Bondar', I.A.; Vinogradova, N.V.; Dem'yanets, L.N.

    1983-01-01

    From the viewpoint of structural chemistry and general regularities controlling formation reactions of compounds and phases in melts, solid and gaseous states, recent achievements in the chemistry of rare earth germanates are generalized. Methods of synthesizing germanates, systems on the base of germanium oxides and rare earths are considered. The data on crystallochemical characteristics are tabulated. Individual compounds of scandium germanate are also characterized. Processes of germanate formation using the data of IR-spectroscopy, X-ray phase analysis are studied. The structure and morphotropic series of rare earth germanates and silicates are determined. Fields of their present and possible future application are considered

  14. CONTAIN code calculations of the effects on the source term of CsI to I/sub 2/ conversion due to severe hydrogen burns

    International Nuclear Information System (INIS)

    Valdez, G.D.; Williams, D.C.

    1986-01-01

    In experiments conducted at Sandia National Laboratories large amounts of elemental iodine were produced when CsI-Al 2 O 3 aerosol was exposed to hydrogen/air combustion. To evaluate some of the implications of the iodide conversion (observed to occur with up to 75% efficiency) for the severe accident source term, computational simulations of representative accident sequences were conducted with the CONTAIN code. The following conclusions can be drawn from this preliminary source term assessment: (1) If the containment sprays are inoperative during the accident, or failed by the hydrogen burn, the late-time source term is almost tripled when the iodide is converted to I 2 . (2) With the sprays active, the amount released without conversion of the CsI aerosol is 63% higher than for the case when conversion occurs. (3) For the case where CsI is converted to I 2 continued operation of the sprays reduces the release by a factor of 40, relative to the case in which the sprays fail at the time of the hydrogen burn. When there is no conversion, the reduction factor for continued spray operation is about a factor of 9, relative to the failed spray case

  15. Toric Varieties and Codes, Error-correcting Codes, Quantum Codes, Secret Sharing and Decoding

    DEFF Research Database (Denmark)

    Hansen, Johan Peder

    We present toric varieties and associated toric codes and their decoding. Toric codes are applied to construct Linear Secret Sharing Schemes (LSSS) with strong multiplication by the Massey construction. Asymmetric Quantum Codes are obtained from toric codes by the A.R. Calderbank P.W. Shor and A.......M. Steane construction of stabilizer codes (CSS) from linear codes containing their dual codes....

  16. Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code; Analisis de un accidente LOCA en contencion de un reactor PWR-W con el codigo GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Perianez Alvarez, V.

    2013-07-01

    The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

  17. Experience and results of MELCOR application for German PWRs

    International Nuclear Information System (INIS)

    Sonnenkalb, M.

    1999-01-01

    An introduction into severe accident research work performed at GRS with regard to the use of the MELCOR code is given in Chapter One of the paper. Experience in applying MELCOR 1.8.3 for German PWRs and results of MELCOR calculations done within the project 'Accident management - Mitigation' for German LWRs are presented in Chapter Two. This 3-year project was finished February 1998. It was funded by the German Ministry for Environment, Nature Conservation and Nuclear Safety - BMU. In Chapter Three, a short overview of a training course on 'Phenomenology of Severe Accidents in PWR-Plants' is given. Mainly due to the interest from German NPPs GRS developed this special training session in 1996. Since 1996 it has been held several times for operators, shift personnel and the management board of two different German NPPs and for lecture of the German NPP training centre in Essen. In Chapter Four, results of the application of MELCOR 1.8.4 for German PWRs are presented. This work is done within a new project on 'Accident Management - Mitigation' for German LWRs. It was started in March 1998 and is again funded by the German Federal Ministry BMU. An objective of this project is to perform further MELCOR calculations, to be used within a PSA level 2 study for a German PWR, which is done at GRS in parallel. The experience of using MELCOR for German PWRs are summarised in Chapter Five. (author)

  18. Hypothesis of Lithocoding: Origin of the Genetic Code as a "Double Jigsaw Puzzle" of Nucleobase-Containing Molecules and Amino Acids Assembled by Sequential Filling of Apatite Mineral Cellules.

    Science.gov (United States)

    Skoblikow, Nikolai E; Zimin, Andrei A

    2016-05-01

    The hypothesis of direct coding, assuming the direct contact of pairs of coding molecules with amino acid side chains in hollow unit cells (cellules) of a regular crystal-structure mineral is proposed. The coding nucleobase-containing molecules in each cellule (named "lithocodon") partially shield each other; the remaining free space determines the stereochemical character of the filling side chain. Apatite-group minerals are considered as the most preferable for this type of coding (named "lithocoding"). A scheme of the cellule with certain stereometric parameters, providing for the isomeric selection of contacting molecules is proposed. We modelled the filling of cellules with molecules involved in direct coding, with the possibility of coding by their single combination for a group of stereochemically similar amino acids. The regular ordered arrangement of cellules enables the polymerization of amino acids and nucleobase-containing molecules in the same direction (named "lithotranslation") preventing the shift of coding. A table of the presumed "LithoCode" (possible and optimal lithocodon assignments for abiogenically synthesized α-amino acids involved in lithocoding and lithotranslation) is proposed. The magmatic nature of the mineral, abiogenic synthesis of organic molecules and polymerization events are considered within the framework of the proposed "volcanic scenario".

  19. 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

  20. RBMK-LOCA-Analyses with the ATHLET-Code

    Energy Technology Data Exchange (ETDEWEB)

    Petry, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Kurfuerstendamm, Berlin (Germany); Domoradov, A.; Finjakin, A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    1995-09-01

    The scientific technical cooperation between Germany and Russia includes the area of adaptation of several German codes for the Russian-designed RBMK-reactor. One point of this cooperation is the adaptation of the Thermal-Hydraulic code ATHLET (Analyses of the Thermal-Hydraulics of LEaks and Transients), for RBMK-specific safety problems. This paper contains a short description of a RBMK-1000 reactor circuit. Furthermore, the main features of the thermal-hydraulic code ATHLET are presented. The main assumptions for the ATHLET-RBMK model are discussed. As an example for the application, the results of test calculations concerning a guillotine type rupture of a distribution group header are presented and discussed, and the general analysis conditions are described. A comparison with corresponding RELAP-calculations is given. This paper gives an overview on some problems posed and experience by application of Western best-estimate codes for RBMK-calculations.

  1. CONDOS: a model and computer code to estimate population and individual radiation doses to man from the distribution, use, and disposal of consumer products that contain radioactive materials

    International Nuclear Information System (INIS)

    O'Donnell, F.R.; McKay, L.R.; Burke, O.W.; Clark, F.H.

    1975-05-01

    A model and computer code (CONDOS) are described that estimate radiation doses to man from distribution, use, and disposal of a variety of consumer products that contain radioactive materials. CONDOS utilizes a generalized format in which the life span of a consumer product is divided into five main stages (distribution, transport, use, disposal, and emergencies) that require descriptions of the activities by which man may be exposed to the product (events) during each stage. These descriptions identify homogeneous groups of exposed persons and, thus, facilitate the selection of individuals who represent the exposed groups. The radiation doses associated with one year of product use to the total body and to selected reference organs and tissues of representative individuals can be estimated for each mode of exposure that is applicable to each event. Summation of the doses to representative individuals yields group and total population doses. An example of the use of CONDOS is given; the radiation doses associated with a hypothetical product are estimated for assumed conditions of exposure. (U.S.)

  2. Maury Journals - German Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — German vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  3. AECL international standard problem ISP-41 FU/1 follow-up exercise (Phase 1): Containment Iodine Computer Code Exercise: Parametric Studies

    International Nuclear Information System (INIS)

    Ball, J.; Glowa, G.; Wren, J.; Ewig, F.; Dickenson, S.; Billarand, Y.; Cantrel, L.; Rydl, A.; Royen, J.

    2001-06-01

    This report describes the results of the second phase of International Standard Problem (ISP) 41, an iodine behaviour code comparison exercise. The first phase of the study, which was based on a simple Radioiodine Test Facility (RTF) experiment, demonstrated that all of the iodine behaviour codes had the capability to reproduce iodine behaviour for a narrow range of conditions (single temperature, no organic impurities, controlled pH steps). The current phase, a parametric study, was designed to evaluate the sensitivity of iodine behaviour codes to boundary conditions such as pH, dose rate, temperature and initial I- concentration. The codes used in this exercise were IODE (IPSN), IODE (NRIR), IMPAIR (GRS), INSPECT (AEAT), IMOD (AECL) and LIRIC (AECL). The parametric study described in this report identified several areas of discrepancy between the various codes. In general, the codes agree regarding qualitative trends, but their predictions regarding the actual amount of volatile iodine varied considerably. The largest source of the discrepancies between code predictions appears to be their different approaches to modelling the formation and destruction of organic iodides. A recommendation arising from this exercise is that an additional code comparison exercise be performed on organic iodide formation, against data obtained from intermediate-scale studies (two RTF (AECL, Canada) and two CAIMAN facility (IPSN, France) experiments have been chosen). This comparison will allow each of the code users to realistically evaluate and improve the organic iodide behaviour sub-models within their codes. (authors)

  4. Containment atmosphere response to external sprays

    Energy Technology Data Exchange (ETDEWEB)

    Green, J.; Almenas, K. [Univ. of Maryland, College Park, MD (United States)

    1995-09-01

    The application of external sprays to a containment steel shell can be an effective energy removal method and has been proposed in the passive AP-600 design. Reduction of the steel shell temperature in contact with the containment atmosphere enhances both heat and mass transfer driving forces. Large scale experimental data in this area is scarce, therefore the measurements obtained from the E series tests conducted at the German HDR facility deserve special attention. These long term tests simulated various severe accident conditions, including external spraying of the hemispherical steel shell. This investigation focuses upon the integral response of the HDR containment atmosphere during spray periods and upon methods by which lumped parameter system codes, like CONTAIN, model the underlying condensation phenomena. Increases in spray water flowrates above a minimum value were ineffective at improving containment pressure reduction since the limiting resistance for energy transfer lies in the noncondensable-vapor boundary layer at the inner condensing surface. The spray created an unstable condition by cooling the upper layers of a heated atmosphere and thus inducing global natural circulation flows in the facility and subsequently, abrupt changes in lighter-than-air noncondensable (J{sub 2}/He) concentrations. Modeling results using the CONTAIN code are outlined and code limitations are delineated.

  5. Containment atmosphere response to external sprays

    International Nuclear Information System (INIS)

    Green, J.; Almenas, K.

    1995-01-01

    The application of external sprays to a containment steel shell can be an effective energy removal method and has been proposed in the passive AP-600 design. Reduction of the steel shell temperature in contact with the containment atmosphere enhances both heat and mass transfer driving forces. Large scale experimental data in this area is scarce, therefore the measurements obtained from the E series tests conducted at the German HDR facility deserve special attention. These long term tests simulated various severe accident conditions, including external spraying of the hemispherical steel shell. This investigation focuses upon the integral response of the HDR containment atmosphere during spray periods and upon methods by which lumped parameter system codes, like CONTAIN, model the underlying condensation phenomena. Increases in spray water flowrates above a minimum value were ineffective at improving containment pressure reduction since the limiting resistance for energy transfer lies in the noncondensable-vapor boundary layer at the inner condensing surface. The spray created an unstable condition by cooling the upper layers of a heated atmosphere and thus inducing global natural circulation flows in the facility and subsequently, abrupt changes in lighter-than-air noncondensable (J 2 /He) concentrations. Modeling results using the CONTAIN code are outlined and code limitations are delineated

  6. SEVERO code - user's manual

    International Nuclear Information System (INIS)

    Sacramento, A.M. do.

    1989-01-01

    This user's manual contains all the necessary information concerning the use of SEVERO code. This computer code is related to the statistics of extremes = extreme winds, extreme precipitation and flooding hazard risk analysis. (A.C.A.S.)

  7. Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics

    International Nuclear Information System (INIS)

    Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

    2013-01-01

    In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

  8. Health Information in German (Deutsch)

    Science.gov (United States)

    ... Tools You Are Here: Home → Multiple Languages → German (Deutsch) URL of this page: https://medlineplus.gov/languages/german.html Health Information in German (Deutsch) To use the sharing features on this page, ...

  9. Expression of the Long Non-Coding RNA HOTAIR Correlates with Disease Progression in Bladder Cancer and Is Contained in Bladder Cancer Patient Urinary Exosomes.

    Directory of Open Access Journals (Sweden)

    Claudia Berrondo

    Full Text Available Exosomes are 30-150nM membrane-bound secreted vesicles that are readily isolated from biological fluids such as urine (UEs. Exosomes contain proteins, micro RNA (miRNA, messenger RNA (mRNA, and long non-coding RNA (lncRNA from their cells of origin. Although miRNA, protein and lncRNA have been isolated from serum as potential biomarkers for benign and malignant disease, it is unknown if lncRNAs in UEs from urothelial bladder cancer (UBC patients can serve as biomarkers. lncRNAs are > 200 nucleotide long transcripts that do not encode protein and play critical roles in tumor biology. As the number of recognized tumor-associated lncRNAs continues to increase, there is a parallel need to include lncRNAs into biomarker discovery and therapeutic target algorithms. The lncRNA HOX transcript antisense RNA (HOTAIR has been shown to facilitate tumor initiation and progression and is associated with poor prognosis in several cancers. The importance of HOTAIR in cancer biology has sparked interest in using HOTAIR as a biomarker and potential therapeutic target. Here we show HOTAIR and several tumor-associated lncRNAs are enriched in UEs from UBC patients with high-grade muscle-invasive disease (HGMI pT2-pT4. Knockdown of HOTAIR in UBC cell lines reduces in vitro migration and invasion. Importantly, loss of HOTAIR expression in UBC cell lines alters expression of epithelial-to-mesenchyme transition (EMT genes including SNAI1, TWIST1, ZEB1, ZO1, MMP1 LAMB3, and LAMC2. Finally, we used RNA-sequencing to identify four additional lncRNAs enriched in UBC patient UEs. These data, suggest that UE-derived lncRNA may potentially serve as biomarkers and therapeutic targets.

  10. QR Codes 101

    Science.gov (United States)

    Crompton, Helen; LaFrance, Jason; van 't Hooft, Mark

    2012-01-01

    A QR (quick-response) code is a two-dimensional scannable code, similar in function to a traditional bar code that one might find on a product at the supermarket. The main difference between the two is that, while a traditional bar code can hold a maximum of only 20 digits, a QR code can hold up to 7,089 characters, so it can contain much more…

  11. [Adjustment of the German DRG system in 2009].

    Science.gov (United States)

    Wenke, A; Franz, D; Pühse, G; Volkmer, B; Roeder, N

    2009-07-01

    The 2009 version of the German DRG system brought significant changes for urology concerning coding of diagnoses, medical procedures and the DRG structure. In view of the political situation and considerable economic pressure, a critical analysis of the 2009 German DRG system is warranted. Analysis of relevant diagnoses, medical procedures and G-DRGs in the versions 2008 and 2009 based on the publications of the German DRG-institute (InEK) and the German Institute of Medical Documentation and Information (DIMDI). The relevant diagnoses, medical procedures and German DRGs in the versions 2008 and 2009 were analysed based on the publications of the German DRG Institute (InEK) and the German Institute of Medical Documentation and Information (DIMDI). Changes for 2009 focus on the development of the DRG structure, DRG validation and codes for medical procedures to be used for very complex cases. The outcome of these changes for German hospitals may vary depending in the range of activities. The German DRG system again gained complexity. High demands are made on correct and complete coding of complex urology cases. The quality of case allocation in the German DRG system was improved. On the one hand some of the old problems (e.g. enterostomata) still persist, while on the other hand new problems evolved out of the attempt to improve the case allocation of highly complex and expensive cases. Time will tell whether the increase in highly specialized DRG with low case numbers will continue to endure and reach acceptable rates of annual fluctuations.

  12. Steam generators. English-German, German-English. Dampferzeuger. Englisch-Deutsch, Deutsch-Englisch

    Energy Technology Data Exchange (ETDEWEB)

    Junge, H D

    1986-01-01

    This pocket dictionary contains the most important technical terms relating to steam generators both in English-German and German-English. Part of the terms go with additional definitions or explanations. Furthermore numerous examples are presented to explain the underlying rules for the formation of word combinations. In addition, entries include a number of general terms, as experience shows that suitable equivalents for use in technical texts are often needed precisely by the specialist. (HAG).

  13. The Use of Film in Teaching German Culture

    Science.gov (United States)

    Figge, Richard C.

    1977-01-01

    Some of the possibilities of teaching German culture through the medium of the fictional film are suggested. Brief descriptions are provided of German films found useful in communicating some aspect or problem of twentieth-century culture. A select bibliography of works containing extensive analyses and interpretations is provided. (SW)

  14. Word order in the Germanic languages

    DEFF Research Database (Denmark)

    Holmberg, Anders; Rijkhoff, Jan

    1998-01-01

    The Germanic branch of Indo-European consists of three main groups (Ruhlen 1987: 327):- East Germanic: Gothic, Vandalic, Burgundian (all extinct);- North Germanic (or: Scandinavian): Runic (extinct), Danish, Swedish, Norwegian, Icelandic, Faroese;- West Germanic: German, Yiddish, Luxembourgeois, ...

  15. 63rd German radiological congress

    International Nuclear Information System (INIS)

    1982-01-01

    The book of abstracts contains abstracts of 171 papers read at the German Radiological Congress in Berlin as well as abstracts of two papers not read for lack of time. Further, there are 31 brief descriptions of the scientific exhibition. Subjects: Diagnosis of gall bladder diseases and inflammatory diseases of the large intestine; hyperthermia and irradiation in tumour therapy; nuclear methods in the diagnosis of growing and displacing processes, skeletal diseases, thromboses, embolisms, gastrointestinal and liver affections; new techniques and methods, diagnostics of the spinal tract; radiooncology; carcinoma of the ovaries; diagnostics and therapy of tumours of the lungs; computerized tomography; angiography; ultrasonic diagnosis. (MG) [de

  16. German Business in Russia

    Directory of Open Access Journals (Sweden)

    Irakliy D. Gvazava

    2013-01-01

    Full Text Available Since Perestroika German-Russian relationships have been steadily developing fueled by close contacts between the leaders of both countries. Boris Yeltsin and Helmut Kohl, Vladimir Putin and Gerhard Schröder, Dmitry Medvedev and Angela Merkel had friendly relations resulted in some fruitful business projects, intergovernmental economic forums etc. In my article I will consider the activities of German companies in Russia, advantages, barriers and expectations

  17. Dictionary of chemistry. English/German

    International Nuclear Information System (INIS)

    Wenske, G.

    1992-01-01

    This English/German dictionary covers more than 100.000 terms from chemistry, chemical engineering and related fields. It also contains molecular formulas, as well as numerous synonyms and areas of application. IUPAC terminology is emphasized, and outdated or rare terminology is indicated. (MM) [de

  18. Attitudes of German Student Teachers on Inclusion

    Science.gov (United States)

    Baar, Robert

    2016-01-01

    The contribution discusses attitudes of German Teacher Training Students on Inclusion based on an empirical analysis containing three elements: Evaluation of students' written exams, results of a survey with closed as open questions and the interpretation of group discussions among students about inclusion. One can see that, though the found-out…

  19. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    Directory of Open Access Journals (Sweden)

    A. Schwenk-Ferrero

    2013-01-01

    Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

  20. Recommendations for the classification of HIV associated neuromanifestations in the German DRG system.

    Science.gov (United States)

    Evers, Stefan; Fiori, W; Brockmeyer, N; Arendt, G; Husstedt, I-W

    2005-09-12

    HIV associated neuromanifestations are of growing importance in the in-patient treatment of HIV infected patients. In Germany, all in-patients have to be coded according to the ICD-10 classification and the German DRG-system. We present recommendations how to code the different primary and secondary neuromanifestations of HIV infection. These recommendations are based on the commentary of the German DRG procedures and are aimed to establish uniform coding of neuromanifestations.

  1. Fifth German-American Frontiers of Engineering Symposium

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2002-05-01

    The agenda book for the Fifth German-American Frontiers of Engineering Symposium contains abstracts of the 16 presentations as well as information on the program, bios of the speakers, contact information for all attendees, and background on the activity.

  2. Intensified colonisation screening according to the recommendations of the German Commission for Hospital Hygiene and Infectious Diseases Prevention (KRINKO): identification and containment of a Serratia marcescens outbreak in the neonatal intensive care unit, Jena, Germany, 2013-2014.

    Science.gov (United States)

    Dawczynski, Kristin; Proquitté, Hans; Roedel, Jürgen; Edel, Brigit; Pfeifer, Yvonne; Hoyer, Heike; Dobermann, Helke; Hagel, Stefan; Pletz, Mathias W

    2016-12-01

    In 2013, the German Commission for Hospital Hygiene and Infectious Disease Prevention (KRINKO) stated that extending weekly colonisation screening from very low birth weight (VLBW) infants (Serratia marcescens. Strains were typed by Pulsed Field Gel Electrophoresis (PFGE). Over 6 months, 19 out of 159 infants acquired S. marcescens. Twelve of the nineteen patients with S. marcescens were non-VLBW infants, and they were colonised significantly earlier than were VLBW infants (median 17 vs. 28 days; p marcescens.

  3. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  4. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  5. Further Generalisations of Twisted Gabidulin Codes

    DEFF Research Database (Denmark)

    Puchinger, Sven; Rosenkilde, Johan Sebastian Heesemann; Sheekey, John

    2017-01-01

    We present a new family of maximum rank distance (MRD) codes. The new class contains codes that are neither equivalent to a generalised Gabidulin nor to a twisted Gabidulin code, the only two known general constructions of linear MRD codes.......We present a new family of maximum rank distance (MRD) codes. The new class contains codes that are neither equivalent to a generalised Gabidulin nor to a twisted Gabidulin code, the only two known general constructions of linear MRD codes....

  6. German energy market 2016

    International Nuclear Information System (INIS)

    Schiffer, Hans-Wilhelm; Weltenergierat, Berlin

    2017-01-01

    The basic orientation of the German energy supply to the increased use of renewable energies, while increasing energy efficiency, is prediscribed by the German government's energy concept and determines the market development. A current overview of the German energy market is given, which provides also this year a concentrated Compilation of the key data of the energy industry. As in the years before, the article not only summarizes general facts about the energy mix, but also goes into detail on the development of the individual energy sources, petroleum, natural gas, brown coal and hard coal, electricity as well as renewable energies. Furthermore, the price trends of international markets and in the domestic market are explained. A current overview of the development of greenhouse gas emissions concludes the contribution. [de

  7. German Idealism Today

    DEFF Research Database (Denmark)

    This collection of essays provides an exemplary overwiew of the diversity and relevance of current scholarship on German Idealism. The importance of German Idealism for contemporary philosophy has recieved growing attention and acknowledgment throughout competing fields of contemporary philosophy...... scholarly debates beyond merely antiquarian perspectives. This renaissance has been a major factor of current efforts to bridge the gap between so-called "nalytic" and so-called "continental" philosophy. The volume provides a selection of readings that contributes to systematic treatments of philosophical...

  8. Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

    Energy Technology Data Exchange (ETDEWEB)

    Bocanegra, R.; Jimenez, G.

    2013-07-01

    The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

  9. [Orthopedic and trauma surgery in the German DRG System 2007].

    Science.gov (United States)

    Franz, D; Kaufmann, M; Siebert, C H; Windolf, J; Roeder, N

    2007-03-01

    The German Diagnosis-Related Groups (DRG) System was further developed into its 2007 version. For orthopedic and trauma surgery, significant changes were made in terms of the coding of diagnoses and medical procedures, as well as in the DRG structure itself. The German Societies for Trauma Surgery and for Orthopedics and Orthopedic Surgery (Deutsch Gesellschaft für Unfallchirurgie, DGU; and Deutsche Gesellschaft für Orthopädie und Orthopädische Chirurgie, DGOOC) once again cooperated constructively with the German DRG Institute InEK. Among other innovations, new International Classification of Diseases (ICD) codes for second-degree burns were implemented. Procedure codes for joint operations, endoprosthetic-surgery and spine surgery were restructured. Furthermore, a specific code for septic surgery was introduced in 2007. In addition, the DRG structure was improved. Case allocation of patients with more than one significant operation was established. Further DRG subdivisions were established according to the patients age and the Patient Clinical Complexity Level (PCCL). DRG developments for 2007 have improved appropriate case allocation, but once again increased the system's complexity. Clinicians need an ever growing amount of specific coding know-how. Still, further adjustments to the German DRG system are required to allow for a correct allocation of cases and funds.

  10. DEMorphy, German Language Morphological Analyzer

    OpenAIRE

    Altinok, Duygu

    2018-01-01

    DEMorphy is a morphological analyzer for German. It is built onto large, compactified lexicons from German Morphological Dictionary. A guesser based on German declension suffixed is also provided. For German, we provided a state-of-art morphological analyzer. DEMorphy is implemented in Python with ease of usability and accompanying documentation. The package is suitable for both academic and commercial purposes wit a permissive licence.

  11. On German Unity 1

    African Journals Online (AJOL)

    German Democratic Republic (GDR) acceded to the Federal Republic of .... living and the shortage of foreign exchange forced the government of the .... manded a great deal of empathy and care above and beyond the normal call of duty. ... The periods of service completed by conscripts in the NPA were set off against the.

  12. Storytelling and German Culture.

    Science.gov (United States)

    Cooper, Connie S. Eigenmann

    The genre of fairytales, one structured form of storytelling, has been labeled "Marchen." German culture is orally transmitted in this generic form, and can be traced to a collection of 210 fairytales, the Grimm brothers'"Kinder-und Taus-Marchen," first published shortly after 1800. For this study, research questions were posed…

  13. Music to Teach German By.

    Science.gov (United States)

    Schulte, Leo

    1985-01-01

    Discusses how music can be intergrated with regular lesson plans to teach German vocabulary, grammar, and history and to give insights into German culture. Also included are sources for basic background information, a list of recordings of the German music, and notes on selecting and presenting it in the language class. (SED)

  14. Proceedings of the EMU Conference on Foreign Languages for Business and the Professions (Dearborn, Michigan, April 5-7, 1984). Part VII: German for Business and the Professions.

    Science.gov (United States)

    Voght, Geoffrey M., Ed.

    Part VII of the proceedings contains five presentations. They are: "German for the Professions: Specialized German for Engineering and the Sciences" (Hannelore Lehr); "German for Business and Economics: A Three-Level Program at Georgetown University" (Barbara Z. Harding); "German for Business and Economics: Criteria for Selection of Specialized…

  15. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  16. CONTAIN independent peer review

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E. [Los Alamos National Lab., NM (United States); Corradini, M.L. [Univ. of Wisconsin, Madison, WI (United States). Nuclear Engineering Dept.; Denning, R.S. [Battelle Memorial Inst., Columbus, OH (United States); Khatib-Rahbar, M. [Energy Research Inc., Rockville, MD (United States); Loyalka, S.K. [Univ. of Missouri, Columbia, MO (United States); Smith, P.N. [AEA Technology, Dorchester (United Kingdom). Winfrith Technology Center

    1995-01-01

    The CONTAIN code was developed by Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission (NRC) to provide integrated analyses of containment phenomena. It is used to predict nuclear reactor containment loads, radiological source terms, and associated physical phenomena for a range of accident conditions encompassing both design-basis and severe accidents. The code`s targeted applications include support for containment-related experimental programs, light water and advanced light water reactor plant analysis, and analytical support for resolution of specific technical issues such as direct containment heating. The NRC decided that a broad technical review of the code should be performed by technical experts to determine its overall technical adequacy. For this purpose, a six-member CONTAIN Peer Review Committee was organized and a peer review as conducted. While the review was in progress, the NRC issued a draft ``Revised Severe Accident Code Strategy`` that incorporated revised design objectives and targeted applications for the CONTAIN code. The committee continued its effort to develop findings relative to the original NRC statement of design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications were considered by the Committee in assigning priorities to the Committee`s recommendations. The Committee determined some improvements are warranted and provided recommendations in five code-related areas: (1) documentation, (2) user guidance, (3) modeling capability, (4) code assessment, and (5) technical assessment.

  17. CONTAIN calculations

    International Nuclear Information System (INIS)

    Scholtyssek, W.

    1995-01-01

    In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident 'medium-sized leak in the cold leg', especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

  18. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  19. An algebraic approach to graph codes

    DEFF Research Database (Denmark)

    Pinero, Fernando

    This thesis consists of six chapters. The first chapter, contains a short introduction to coding theory in which we explain the coding theory concepts we use. In the second chapter, we present the required theory for evaluation codes and also give an example of some fundamental codes in coding...... theory as evaluation codes. Chapter three consists of the introduction to graph based codes, such as Tanner codes and graph codes. In Chapter four, we compute the dimension of some graph based codes with a result combining graph based codes and subfield subcodes. Moreover, some codes in chapter four...

  20. High Energy Transport Code HETC

    International Nuclear Information System (INIS)

    Gabriel, T.A.

    1985-09-01

    The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs

  1. Consulting-Research Froblems with German and American Multinational Firms.

    Science.gov (United States)

    Hildebrandt, Herbert W.

    International researchers need to be aware of international problems and multinational managerial codes when they work with worldwide organizations. This paper develops the premise that consulting with German multinational companies is more complex than consulting with or researching for American firms. Discussion focuses on the following three…

  2. Brazilian Portuguese and German in contact in two virtual communities

    Directory of Open Access Journals (Sweden)

    Layla Cristina Iapechino Souto

    2017-01-01

    Full Text Available This paper presents an analysis of code-switching between Brazilian Portuguese and German language in two virtual communities on Facebook: Brasileiros em Berlim and Brasileiros e Brasileiras em Berlim. We have adopted the concepts of durability, permeability and liminality traced by Zinkhahn-Rhobodes (2015 to observe the permeability of the linguistic border between these two languages. 

  3. CONTAIN calculations; CONTAIN-Rechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Scholtyssek, W.

    1995-08-01

    In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident `medium-sized leak in the cold leg`, especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

  4. School of German Language

    Directory of Open Access Journals (Sweden)

    Sergei V. Evteev

    2014-01-01

    Full Text Available Department of German is one of the oldest language departments at MGIMO. Since its foundation in 1944 the military experienced teachers of the department, most of whom were native speakers, have begun to develop a unique method of teaching the German language, thereby revolutionize learning this foreign language. The first steps made under the supervision of the Department of Antonina V. Celica. The department refused to conventional time and is still used in universities such as the Moscow Linguistic University, separate teaching phonetics, grammar and vocabulary, which was due to the specific objectives set for the teaching staff: prepare for short term specialists in international relations, active Germanspeaking. The department can be proud of its graduates, many of whom continue his career in the walls of native high school. Many graduates have dedicated their lives to serving the State in the Ministry of Foreign Affairs.

  5. DLLExternalCode

    Energy Technology Data Exchange (ETDEWEB)

    2014-05-14

    DLLExternalCode is the a general dynamic-link library (DLL) interface for linking GoldSim (www.goldsim.com) with external codes. The overall concept is to use GoldSim as top level modeling software with interfaces to external codes for specific calculations. The DLLExternalCode DLL that performs the linking function is designed to take a list of code inputs from GoldSim, create an input file for the external application, run the external code, and return a list of outputs, read from files created by the external application, back to GoldSim. Instructions for creating the input file, running the external code, and reading the output are contained in an instructions file that is read and interpreted by the DLL.

  6. Error Correcting Codes

    Indian Academy of Sciences (India)

    Science and Automation at ... the Reed-Solomon code contained 223 bytes of data, (a byte ... then you have a data storage system with error correction, that ..... practical codes, storing such a table is infeasible, as it is generally too large.

  7. In silico comparison of genomic regions containing genes coding for enzymes and transcription factors for the phenylpropanoid pathway in Phaseolus vulgaris L. and Glycine max L. Merr

    Directory of Open Access Journals (Sweden)

    Yarmilla eReinprecht

    2013-09-01

    Full Text Available Legumes contain a variety of phytochemicals derived from the phenylpropanoid pathway that have important effects on human health as well as seed coat color, plant disease resistance and nodulation. However, the information about the genes involved in this important pathway is fragmentary in common bean (Phaseolus vulgaris L.. The objectives of this research were to isolate genes that function in and control the phenylpropanoid pathway in common bean, determine their genomic locations in silico in common bean and soybean, and analyze sequences of the 4CL gene family in two common bean genotypes. Sequences of phenylpropanoid pathway genes available for common bean or other plant species were aligned, and the conserved regions were used to design sequence-specific primers. The PCR products were cloned and sequenced and the gene sequences along with common bean gene-based (g markers were BLASTed against the Glycine max v.1.0 genome and the P. vulgaris v.1.0 (Andean early release genome. In addition, gene sequences were BLASTed against the OAC Rex (Mesoamerican genome sequence assembly. In total, fragments of 46 structural and regulatory phenylpropanoid pathway genes were characterized in this way and placed in silico on common bean and soybean sequence maps. The maps contain over 250 common bean g and SSR (simple sequence repeat markers and identify the positions of more than 60 additional phenylpropanoid pathway gene sequences, plus the putative locations of seed coat color genes. The majority of cloned phenylpropanoid pathway gene sequences were mapped to one location in the common bean genome but had two positions in soybean. The comparison of the genomic maps confirmed previous studies, which show that common bean and soybean share genomic regions, including those containing phenylpropanoid pathway gene sequences, with conserved synteny. Indels identified in the comparison of Andean and Mesoamerican common bean sequences might be used to develop

  8. Nonstationary pressure build up in full-pressure containments after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1977-01-01

    The time histories of pressure, temperature and pressure difference during the pressure build up phase of a loss-of-coolant accident (LOCA) in the primary system in full-pressure containments of water cooled nuclear power reactors are treated. These are important for the design of such containments. The experiments within the German research program RS 50 ''Druckverteilung im Containment'' offered, for the first time, the opportunity to observe experimentally fluid-dynamic processes in a multiple divided full-pressure containment, and to test at the same time, computer codes which serve to describe the physical processes during the LOCA. The comparison of the results calculated by the computer codes ZOCO VI and DDIFF with the experimental results showed apparent deviations by special arrangements of the compartments and the vent flow paths of a model containment for the calculation of time dependent pressure-, temperature- and pressure difference-histories. The deviations lead to the development of the analytical model and computer code COFLOW. This new model was primarily designed to deal with the fluid-dynamic processes in the beginning phase of the blowdown as maximal pressure differences appear. Furthermore, it can be used to determine the maximum containment pressure, as well as for long term calculations. The analytical model and computer code COFLOW shows a better correlation between theory and experiment than previous codes

  9. CONTAIN independent peer review

    International Nuclear Information System (INIS)

    Boyack, B.E.; Corradini, M.L.; Khatib-Rahbar, M.; Loyalka, S.K.; Smith, P.N.

    1995-01-01

    The CONTAIN code was developed by Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission (NRC) to provide integrated analyses of containment phenomena. It is used to predict nuclear reactor containment loads, radiological source terms, and associated physical phenomena for a range of accident conditions encompassing both design-basis and severe accidents. The code's targeted applications include support for containment-related experimental programs, light water and advanced light water reactor plant analysis, and analytical support for resolution of specific technical issues such as direct containment heating. The NRC decided that a broad technical review of the code should be performed by technical experts to determine its overall technical adequacy. For this purpose, a six-member CONTAIN Peer Review Committee was organized and a peer review as conducted. While the review was in progress, the NRC issued a draft ''Revised Severe Accident Code Strategy'' that incorporated revised design objectives and targeted applications for the CONTAIN code. The committee continued its effort to develop findings relative to the original NRC statement of design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications were considered by the Committee in assigning priorities to the Committee's recommendations. The Committee determined some improvements are warranted and provided recommendations in five code-related areas: (1) documentation, (2) user guidance, (3) modeling capability, (4) code assessment, and (5) technical assessment

  10. [German ophthalmologists and NSDAP

    Science.gov (United States)

    Rohrbach, Jens Martin

    2008-01-01

    Approximately 40-45 % of all German physicians joined the National Socialist German Workers Party (NSDAP) until 1945. Reasons for party membership are manifold and still a matter of debate. Very likely, the extraordinary high representation of medical doctors in the NSDAP was rather a result of active entry than recruitment by the party. There are only few data concerning the willingness of ophthalmologists to become a party member ("Parteigenosse", "Pg"). According to the list of University teachers in Germany ("Hochschullehrerkarte"; Federal Archive, Berlin), the list of the members of the German Ophthalmological Society (DOG) of 1934 and especially the list of NSDAP-members (Federal Archive, Berlin) the following conclusions can be drawn: 1. Directors of German University eye hospitals (chairmen) were members of the NSDAP with a frequency of 23% in 1933 and 48% in 1938 as well as in 1943. The motivation for joining the party was most likely the perspective of acceleration of the academic career. 2. "Only" 30% of the ophthalmologists working in private praxis were "Pg" (until 1945). 3. Both chairmen and ophthalmologists in private praxis were equally hindered to join the NSDAP between May 1st 1933 and May 1st 1937 when the party temporarily stopped registration. 4. The majority of ophthalmologists who joined the NSDAP were born between 1880 and 1900 and thus had taken part in World War I as soldiers or had experienced the times of need after WW I. Only few ophthalmologists succeeded in the NS-hierarchy and probably only one ophthalmologist, Walther Löhlein from Berlin, came in personal contact with Adolf Hitler who was constantly in fear for his sight after his eye injury in October 1918. The "Law for the prevention of genetically disabled offsprings" ("Gesetz zur Verhütung erbkranken Nachwuchses") from July 14th, 1933 separated ophthalmologists into two parties: those advocating sterilization to a high degree and those recommending sterilization only

  11. Modeling of local steam condensation on walls in presence of non-condensable gases. Application to a loca calculation in reactor containment using the multidimensional geyser/tonus code

    Energy Technology Data Exchange (ETDEWEB)

    Benet, L.V.; Caroli, C.; Cornet, P. [Commissariat a l`Energie Atomique, Gif sur Yvette (France)] [and others

    1995-09-01

    This paper reports part of a study of possible severe pressurized water reactor (PWR) accidents. The need for containment modeling, and in particular for a hydrogen risk study, was reinforced in France after 1990, with the requirement that severe accidents must be taken into account in the design of future plants. This new need of assessing the transient local hydrogen concentration led to the development, in the Mechanical Engineering and Technology Department of the French Atomic Energy Commission (CEA/DMT), of the multidimensional code GEYSER/TONUS for containment analysis. A detailed example of the use of this code is presented. The mixture consisted of noncondensable gases (air or air plus hydrogen) and water vapor and liquid water. This is described by a compressible homogeneous two-phase flow model and wall condensation is based on the Chilton-Colburn formula and the analogy between heat and mass transfer. Results are given for a transient two-dimensional axially-symmetric computation for the first hour of a simplified accident sequence. In this there was an initial injection of a large amount of water vapor followed by a smaller amount and by hydrogen injection.

  12. Importance of core/concrete interactions for German risk investigations and experimental verification

    International Nuclear Information System (INIS)

    Rohde, J.; Hicken, E.F.; Friederichs, H.G.; Schroedl, E.

    1987-01-01

    The relevance of Molten Core Concrete Interactions (MCCI) for risk oriented investigations of German LWR-plants is evaluated. The problems of MCCI have been intensely investigated since the mid seventies in connection with the German Risk Study, Phase A and B on PWR plants of German design. Many examinations of both theoretical and experimental nature have led to the development of computer codes like WECHSL. The basis for verification is the internationally well accepted BETA experiment. Code WECHSL and the knowledge gained from the BETA experiment have been applied for the final investigations in German Risk Study, Phase B. Knowledge gained will be illustrated and its importance MCCI for German LWR-concepts will be shown

  13. Marlene Dietrich in the German Classroom: A German Film Project--Humanities through the Golden Age of German Cinema.

    Science.gov (United States)

    Flippo, Hyde

    1993-01-01

    Marlene Dietrich and other classic performers of German cinema can serve to open up a whole new realm for students of German, at secondary and postsecondary levels. By researching and viewing German and American film classics, students have opportunity to learn more about German language and an important element of German culture that has had…

  14. Geographic data: Zip Codes (Shape File)

    Data.gov (United States)

    Montgomery County of Maryland — This dataset contains all zip codes in Montgomery County. Zip codes are the postal delivery areas defined by USPS. Zip codes with mailboxes only are not included. As...

  15. Elaboration of data and documents intended to complement and expand the German series of nuclear engineering codes. 3. Technical report. Non-destructive testing of austenitic welds and claddings

    International Nuclear Information System (INIS)

    Waidele, H.

    1997-01-01

    This 3. technical report presents a literature study on non-destructive testing of austenitic welds and claddings. NDT of claddings was the subject of a previous BMU project report SR 2024, so that this report contains only an update covering the latest developments in this subject area, and NDT of austenitic welds is the major subject of the report in hand. The literature study shows that improvements of ultrasonic test results for austenitic welds are expected to be achieved soon as a result of application of novel testing methods, advanced signal processing algorithms, and reduced anisotropy of austenitic welds due to specific welding techniques. Enhanced information is expected to be achieved from radiography tests through improvements available now, such as digitization of conventional radiographs combined with computer-assisted evaluation methods. As to the inspection of components with wall thickness up to 10 mm, low-frequency methods or eddy current methods will increasingly be applied in future as complementing methods supplying additional information. (orig./CB) [de

  16. Backlash against American psychology: an indigenous reconstruction of the history of German critical psychology.

    Science.gov (United States)

    Teo, Thomas

    2013-02-01

    After suggesting that all psychologies contain indigenous qualities and discussing differences and commonalities between German and North American historiographies of psychology, an indigenous reconstruction of German critical psychology is applied. It is argued that German critical psychology can be understood as a backlash against American psychology, as a response to the Americanization of German psychology after WWII, on the background of the history of German psychology, the academic impact of the Cold War, and the trajectory of personal biographies and institutions. Using an intellectual-historical perspective, it is shown how and which indigenous dimensions played a role in the development of German critical psychology as well as the limitations to such an historical approach. Expanding from German critical psychology, the role of the critique of American psychology in various contexts around the globe is discussed in order to emphasize the relevance of indigenous historical research.

  17. Electricity: the German example

    International Nuclear Information System (INIS)

    Huet, Sylvestre

    2013-01-01

    The author proposes some comments on the content of the Energiewende, i.e. the definition of the energy transition in Germany which aims at producing and consuming a green energy, without carbon nor nuclear. He comments the German energy mix for 2010 in terms of electricity production per origin (nuclear, coal and lignite, gas, oil, wind, solar photovoltaic, other renewable sources) and of installed capacities per origin. He notices that gas and coal still have a major weight in this mix, and discusses the content of a scenario based 100 per cent renewable energies as it has been studied by the Fraunhofer Institute, notably in terms of production level and of costs

  18. [Orthopedic and trauma surgery in the German DRG system 2008].

    Science.gov (United States)

    Franz, D; Kaufmann, M; Siebert, C H; Windolf, J; Roeder, N

    2008-04-01

    The German DRG (diagnosis-related groups) system has been modified and updated into version 2008. For orthopedic and trauma surgery significant changes concerning coding of diagnoses, medical procedures and the DRG structure were made. The modified version has been analyzed in order to ascertain whether the DRG system is suitably qualified to fulfill the demands of the reimbursement system or whether further improvements are necessary. Analysis of the severity of relevant side-effect diagnoses, medical procedures and G-DRGs in the versions 2007 and 2008 was carried out based on the publications of the German DRG institute (InEK) and the German Institute of Medical Documentation and Information (DIMDI). Changes for 2008 focused on the development of DRG structure, DRG validation and codes for medical procedures. The outcome of these changes for German hospitals may vary depending on the range of activities. G-DRG system has become even more complex and the new regulations have also resulted in new problems associated with complications.. High demands are made on correct and complete coding of complex orthopedic and trauma surgery cases. Quality of case allocation within the G-DRG system has been improved. Nevertheless, further improvements of the G-DRG system are necessary, especially for cases with severe injuries.

  19. THAI experimental programme for containment safety assessment under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, S.; Freitag, M. [Becker Technologies GmbH, Eschborn (Germany); Poss, G.

    2016-05-15

    The THAI (THAI = Thermal hydraulics, Hydrogen, Aerosols, Iodine) experimental programme aims to address open questions concerning the behavior of hydrogen, iodine and aerosols in the containment of water cooled reactors. Since its construction in 2000, THAI programme is being performed in the frame of various national projects (sponsored by German Federal Ministry for Economic Affairs and Energy, BMWi) and two international joint projects (under auspices of OECD/NEA). THAI experimental data have been widely used for the validation and further development of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes with 3D capabilities. Selected examples of code benchmark exercises performed based on the THAI data include; hydrogen distribution experiment (ISP-47 and OECD/NEA THAI code benchmark), hydrogen combustion behaviour (ISP-49), hydrogen mitigation by PARs (OECD/NEA THAI-2 code benchmark), iodine/surface interactions, iodine mass transfer, and iodine transport and multi-compartment behaviour (EU-SARNET and EU-SARNET2), thermal-hydraulic tests (German CFD-network). In the present paper, a brief overview on the THAI experiments and their role in the containment safety assessment is discussed.

  20. ASTEC V2. Overview of code development and application at GRS

    International Nuclear Information System (INIS)

    Reinke, N.; Nowack, H.; Sonnenkalb, M.

    2011-01-01

    The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed since 1996 by the French IRSN and the German 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. Thus, the main ASTEC application fields are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses as well as a combination of multiple modules for coupled effects testing and integral analyses. Subject of this paper is an overview of the new V2 series of the ASTEC code system and presentation of exemplary results for the application to severe accidents sequences at PWRs. (orig.)

  1. FAUST/CONTAIN; FAUST/CONTAIN

    Energy Technology Data Exchange (ETDEWEB)

    Cherdron, W.; Minges, J.; Sauter, H.; Schuetz, W.

    1995-08-01

    The FAUNA facility has been restructured after completion of the sodium fire experiments. It is now serving LWR research, cf. report II on program no. 32.21.02 concerning steam explosions. The CONTAIN code system for computing the thermodynamic, aerosol and radiological phenomena in a containment under severe accident conditions is being developed with a new to fission product release and transport. (orig.)

  2. Construction of new quantum MDS codes derived from constacyclic codes

    Science.gov (United States)

    Taneja, Divya; Gupta, Manish; Narula, Rajesh; Bhullar, Jaskaran

    Obtaining quantum maximum distance separable (MDS) codes from dual containing classical constacyclic codes using Hermitian construction have paved a path to undertake the challenges related to such constructions. Using the same technique, some new parameters of quantum MDS codes have been constructed here. One set of parameters obtained in this paper has achieved much larger distance than work done earlier. The remaining constructed parameters of quantum MDS codes have large minimum distance and were not explored yet.

  3. Basic criteria and application examples of German utility PLIM concept

    International Nuclear Information System (INIS)

    Sgarz, G.; Metzner, K.J.

    2002-01-01

    As a consequence of the consensus negotiations between the present Federal German Government and the German utilities the new Atomic Energy Law was set into force in April 2002. The main issues are: 1. Phase out of NPP-operation after a maximum lifetime of 32 years without any claims for compensation. 2. Termination of spent fuel reprocessing and switching over to direct final storage. Stop of spent fuel casks shipment in 2005. 3. Intermediate storage facilities are to be provided on each power plant site. 4. The promotion clause for nuclear energy is cancelled, the construction of new NPP's is prohibited. 5. The NPP safety status has to be kept on a high level standard. A periodic safety assessment must be performed 'according to the state of the art' based on up-to-date codes and standards in a 10-year interval. As a consequence, the future German policies and strategies are based on this law

  4. [Consistency and Reliability of MDK Expertise Examining the Encoding in the German DRG System].

    Science.gov (United States)

    Gaertner, T; Lehr, F; Blum, B; van Essen, J

    2015-09-01

    Hospital inpatient stays are reimbursed on the basis of German diagnosis-related groups (G-DRG). The G-DRG classification system is based on complex coding guidelines. The Medical Review Board of the Statutory Health Insurance Funds (MDK) examines the encoding by hospitals and delivers individual expertises on behalf of the German statutory health insurance companies in cases in which irregularities are suspected. A study was conducted on the inter-rater reliability of the MDK expertises regarding the scope of the assessment. A representative sample of 212 MDK expertises was taken from a selected pool of 1 392 MDK expertises in May 2013. This representative sample underwent a double-examination by 2 independent MDK experts using a special software based on the 3MTM G-DRG Grouper 2013 of 3M Medica, Germany. The following items encoded by the hospitals were examined: DRG, principal diagnosis, secondary diagnoses, procedures and additional payments. It was analysed whether the results of MDK expertises were consistent, reliable and correct. 202 expertises were eligible for evaluation, containing a total of 254 questions regarding one or more of the 5 items encoded by hospitals. The double-examination by 2 independent MDK experts showed matching results in 187 questions (73.6%) meaning they had been examined consistently and correctly. 59 questions (23.2%) did not show matching results, nevertheless they had been examined correctly regarding the scope of the assessment. None of the principal diagnoses was significantly affected by inconsistent or wrong judgment. A representative sample of MDK expertises examining the DRG encoding by hospitals showed a very high percentage of correct examination by the MDK experts. Identical MDK expertises cannot be achieved in all cases due to the scope of the assessment. Further improvement and simplification of codes and coding guidelines are required to reduce the scope of assessment with regard to correct DRG encoding and its

  5. German visits to CERN

    CERN Multimedia

    2007-01-01

    State secretary to Germany's Federal Ministry of Education and Research, Frieder Meyer-Krahmer, with CERN's Director-General Robert Aymar.On 21 February, Professor Frieder Meyer-Krahmer, State Secretary to Germany's Federal Ministry of Education and Research, came to CERN. He visited the ALICE and ATLAS experiments and the computing centre before meeting the CERN's Director-General, some German physicists and members of the top management. The Minister of Science, Research and the Arts of the Baden-Württemberg regional government, Peter Frankenberg, and CERN's Director-General, Robert Aymar, signing an agreement on education. In the background: Sigurd Lettow, CERN's Director of Finance and Human Resources, and Karl-Heinz Meisel, Rector of the Fachhochschule Karlsruhe. The Minister of Science, Research and the Arts of the Baden-Württemberg regional government, Prof. Peter Frankenberg, visited CERN on 23 February. He was accompanied by the Rector of the Fachhochschule Karlsruhe, Prof. Karl-Heinz Meisel, and b...

  6. Investigations for the implementation of catalytic recombiners in large dry containments in Germany

    International Nuclear Information System (INIS)

    Rohde, J.; Tiltmann, M.; Froehmel, T.

    1997-01-01

    During the past few years, several concepts of mitigation have been developed and tested to limit the hydrogen concentrations in the containment atmosphere during the course of a severe accident. Extensive efforts have been given, especially in Germany and Canada, to investigate the use of catalytic recombiners. Based on the outcome of these research efforts in Germany it was recommended by the Reactor Safety Commission (RSK) in June 1994 to implement a hydrogen mitigation system, based on catalytic recombiners in large dry containments of PWR plants. Investigations are carried out at GRS, sponsored by the German Ministry of Environment, Nature Conservation and Nuclear Safety (BMU), to develop basic requirements for the implementation of a catalytic recombiner system in large dry containments. Severe accidents scenarios were calculated with the system code MELCOR to determine the mass- and energy release rates from the primary system into the containment, necessary to prepare the input data for the containment code calculations. A detailed nodalisation of the containment system of a German PWR plant (Konvoi-type) was developed for the code RA-LOC MOD4 to investigate the effectiveness of a catalytic recombiner system which consists of 53 of such devices, being distributed in the complex room arrangement. The effectiveness of such a system is demonstrated by comparing a representative severe accident sequence without and with the catalytic recombination of hydrogen. The results showed, that only in some limited areas in the containment combustible gas mixtures were formed for a limited time span and that at the end of the first day after the onset of the accident the catalytic reaction is limited due to oxygen depletion. The work is still in progress while additional severe accident sequences have to be analyzed to develop some generic guidelines for the implementation of a catalytic recombiner system in large dry containments. (author)

  7. The German reactor safety study

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1980-01-01

    The most important results of the German risk study of a nuclear power plant equipped with a pressurized water reactor were published in August 1979. The main volume of the study with the approach used and the results elaborated has been available for reference since late 1979. Eight technical volumes contain detailed descriptions and documentations of the investigations carried out. The reference facility used as a basis for the technical plant studies was unit B of the Biblis Nuclear Power Station, a KWU PWR of 3750 MW thermal power. This contribution provides more detailed explanations of the methods and the results of the risk study illustrated by examples. The description refers to accident categories and categories of radioactivity releases, probabilities of specific sequences of accident events, and the damage associated with core meltdown accidents as a function of various types of failure. For purposes of evaluation and application of the results the limits in the basic assumptions of the study are referred to. (orig./HP) [de

  8. Becoming German: Integration, Citizenship and Territorialization of Germanness

    DEFF Research Database (Denmark)

    Fogelman, Tatiana

    2017-01-01

    understandings of integration and Germanness, this paper highlights the neglected aspect of the ascendance of Integrationspolitik since the turn of the century: namely how it superseded previous regime of completely bifurcated migration policy for "foreigners" on the one hand, and so-called "settlers" of German......, seen ever more as residing within its state territory rather than some diffuse cultural-linguistic space. Moving our understanding of Germanness beyond the "ethnic nationhood model" (Faist 2008), I argue thus that, in conjunction with the new citizenship law, the emergence of Integrationspolitik...

  9. Dictionary of heat exchanger technology. English-German, German-English. Woerterbuch der Waermeaustauschertechnik. Englisch-Deutsch, Deutsch-Englisch

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, H P [comp.

    1989-01-01

    This dictionary contains more than 6,000 terms and numerous explanations and comprises all types of shell-and-tube and tubular heat exchangers including condensers, feedwater heaters, air heaters, evaporators, vaporizers, steam generators, steam boilers as well as plate-and-frame heat exchangers, cooling towers, and special designs, and the related technical fields such as thermal and mass transfer, thermodynamics, fluids engineering, and strength calculation. Part 1 contains the English-German version, Part 2 the German-English version and Annex 1 the figures for explaining the most important heat exchanger designs. (orig.).

  10. Abortion checks at German-Dutch border.

    Science.gov (United States)

    Von Baross, J

    1991-05-01

    The commentary on West German abortion law, particularly in illegal abortion in the Netherlands, finds the law restrictive and in violation of the dignity and rights of women. The Max-Planck Institute in 1990 published a study that found that a main point of prosecution between 1976 and 1986, as reported by Der Spiegal, was in border crossings from the Netherlands. It is estimated that 10,000 annually have abortions abroad, and 6,000 to 7,000 in the Netherlands. The procedure was for an official to stop a young person and query about drugs; later the woman would admit to an abortion, and be forced into a medical examination. The German Penal Code Section 218 stipulates abortion only for certain reasons testified to by a doctor other than the one performing the abortion. Counseling on available social assistance must be completed 3 days prior to the abortion. Many counseling offices are church related and opposed to abortions. Many doctors refuse legally to certify, and access to abortion is limited. The required hospital stay is 3-4 nights with no day care facilities. Penal Code Section 5 No. 9 allows prosecution for uncounseled illegal abortion. Abortion law reform is anticipated by the end of 1992 in the Bundestag due to the Treaty or the Unification of Germany. The Treaty states that the rights of the unborn child must be protected and that pregnant women relieve their distress in a way compatible with the Constitution, but improved over legal regulations from either West or East Germany, which permits abortion on request within 12 weeks of conception without counseling. It is hoped that the law will be liberalized and Penal Code Section 5 No. 9 will be abolished.

  11. Aerosol in the containment

    International Nuclear Information System (INIS)

    Lanza, S.; Mariotti, P.

    1986-01-01

    The US program LACE (LWR Aerosol Containment Experiments), in which Italy participates together with several European countries, Canada and Japan, aims at evaluating by means of a large scale experimental activity at HEDL the retention in the pipings and primary container of the radioactive aerosol released following severe accidents in light water reactors. At the same time these experiences will make available data through which the codes used to analyse the behaviour of the aerosol in the containment and to verify whether by means of the codes of thermohydraulic computation it is possible to evaluate with sufficient accuracy variable influencing the aerosol behaviour, can be validated. This report shows and compares the results obtained by the participants in the LACE program with the aerosol containment codes NAVA 5 and CONTAIN for the pre-test computations of the test LA 1, in which an accident called containment by pass is simulated

  12. Development of NPP Safety Requirements into Kenya's Grid Codes

    Energy Technology Data Exchange (ETDEWEB)

    Ndirangu, Nguni James; Koo, Chang Choong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand.

  13. Development of NPP Safety Requirements into Kenya's Grid Codes

    International Nuclear Information System (INIS)

    Ndirangu, Nguni James; Koo, Chang Choong

    2015-01-01

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand

  14. How old are Germanic lambs?

    DEFF Research Database (Denmark)

    Vrieland, Seán D.

    2017-01-01

    Gothic and Gutnish lamb with the meaning ‘sheep’ sets these two languages apart from the rest of Germanic, and is the most common piece of evidence used to claim they share a close connection. Yet the same meaning is found in the descendants of Proto-Fennic *lambaz, a loan from Proto-Germanic, an......Gothic and Gutnish lamb with the meaning ‘sheep’ sets these two languages apart from the rest of Germanic, and is the most common piece of evidence used to claim they share a close connection. Yet the same meaning is found in the descendants of Proto-Fennic *lambaz, a loan from Proto...

  15. … but You Are Not German." -- Afro-German Culture and Literature in the German Language Classroom

    Science.gov (United States)

    Schenker, Theresa; Munro, Robert

    2016-01-01

    Units and classes dedicated to multiculturalism in Germany have predominantly focused on Turkish-German literature and culture. Afro-Germans have been a minority whose culture and literature have only marginally been included in German classes, even though Afro-Germans have been a part of Germany for centuries and have undergone efforts at…

  16. Ageing management in German nuclear power plants

    International Nuclear Information System (INIS)

    Becker, D.E.; Reiner, M.

    1998-01-01

    In Germany, the term 'ageing management' comprises several aspects. A demand for a special ageing monitoring programme is not explicitly contained in the regulations. However, from the Atomic Energy Act and its regulations results the operator's obligation to perform extensive measures to maintain the quality of the plant and the operating personnel working in the plant. From this point of view, comprehensive ageing management in German nuclear power plants has taken place right from the start under the generic term of quality assurance. (author)

  17. Generic Containment: Detailed comparison of containment simulations performed on plant scale

    International Nuclear Information System (INIS)

    Kelm, St.; Klauck, M.; Beck, S.; Allelein, H.-J.; Preusser, G.; Sangiorgi, M.; Klein-Hessling, W.; Bakalov, I.; Bleyer, A.; Bentaib, A.; Kljenak, I.; Stempniewicz, M.; Kostka, P.; Morandi, S.; Ada del Corno, B.; Bratfisch, C.; Risken, T.; Denk, L.; Parduba, Z.; Paci, S.

    2014-01-01

    Highlights: • Consequent implementation of the recommendations derived from the OECD/NEA ISP-47. • Phenomenological code-to-code comparison performed on plant scale. • Systematic identification and elimination of the user effect. • Identification of fundamental differences in the model basis. • Application to PAR system analysis. - Abstract: One outcome of the OECD/NEA ISP-47 activity was the recommendation to elaborate a ‘Generic Containment’ in order to allow comparing and rating the results obtained by different lumped-parameter models on plant scale. Within the European SARNET2 project ( (http://www.sar-net.eu)), such a Generic Containment nodalisation, based on a German PWR (1300 MW el ), was defined. This agreement on the nodalisation allows investigating the remaining differences among the results, especially the ‘user-effect’, related to the modelling choices, as well as fundamental differences in the underlying model basis in detail. The methodology applied in order to compare the different code predictions consisted of a series of three benchmark steps with increasing complexity as well as a systematic comparison of characteristic variables and observations. This paper summarises the benchmark series, the lessons learned during specifying the steps, comparing and discussing the results and finally gives an outlook on future steps

  18. Synthesizing Certified Code

    Science.gov (United States)

    Whalen, Michael; Schumann, Johann; Fischer, Bernd

    2002-01-01

    Code certification is a lightweight approach to demonstrate software quality on a formal level. Its basic idea is to require producers to provide formal proofs that their code satisfies certain quality properties. These proofs serve as certificates which can be checked independently. Since code certification uses the same underlying technology as program verification, it also requires many detailed annotations (e.g., loop invariants) to make the proofs possible. However, manually adding theses annotations to the code is time-consuming and error-prone. We address this problem by combining code certification with automatic program synthesis. We propose an approach to generate simultaneously, from a high-level specification, code and all annotations required to certify generated code. Here, we describe a certification extension of AUTOBAYES, a synthesis tool which automatically generates complex data analysis programs from compact specifications. AUTOBAYES contains sufficient high-level domain knowledge to generate detailed annotations. This allows us to use a general-purpose verification condition generator to produce a set of proof obligations in first-order logic. The obligations are then discharged using the automated theorem E-SETHEO. We demonstrate our approach by certifying operator safety for a generated iterative data classification program without manual annotation of the code.

  19. RFQ simulation code

    International Nuclear Information System (INIS)

    Lysenko, W.P.

    1984-04-01

    We have developed the RFQLIB simulation system to provide a means to systematically generate the new versions of radio-frequency quadrupole (RFQ) linac simulation codes that are required by the constantly changing needs of a research environment. This integrated system simplifies keeping track of the various versions of the simulation code and makes it practical to maintain complete and up-to-date documentation. In this scheme, there is a certain standard version of the simulation code that forms a library upon which new versions are built. To generate a new version of the simulation code, the routines to be modified or added are appended to a standard command file, which contains the commands to compile the new routines and link them to the routines in the library. The library itself is rarely changed. Whenever the library is modified, however, this modification is seen by all versions of the simulation code, which actually exist as different versions of the command file. All code is written according to the rules of structured programming. Modularity is enforced by not using COMMON statements, simplifying the relation of the data flow to a hierarchy diagram. Simulation results are similar to those of the PARMTEQ code, as expected, because of the similar physical model. Different capabilities, such as those for generating beams matched in detail to the structure, are available in the new code for help in testing new ideas in designing RFQ linacs

  20. German concept and status of the disposal of spent fuel elements from German research reactors

    International Nuclear Information System (INIS)

    Komorowski, K.; Storch, S.; Thamm, G.

    1995-01-01

    Eight research reactors with a power ≥ 100 kW are currently being operated in the Federal Republic of Germany. These comprise three TRIGA-type reactors (power 100 kW to 250 kW), four swimming-pool reactors (power 1 MW to 10 MW) and one DIDO type reactor (power 23 MW). The German research reactors are used for neutron scattering for basic research in the field of solid state research, neutron metrology, for the fabrication of isotopes and for neutron activation analysis for medicine and biology, for investigating the influence of radiation on materials and for nuclear fuel behavior. It will be vital to continue current investigations in the future. Further operation of the German research reactors is therefore indispensable. Safe, regular disposal of the irradiated fuel elements arising now and in future operation is of primary importance. Furthermore, there are several plants with considerable quantities of spent fuel, the safe disposal of which is a matter of urgency. These include above all the VKTA facilities in Rossendorf and also the TRIGA reactors, where disposal will only be necessary upon decommissioning. The present paper report is concerned with the disposal of fuel from the German research reactors. It briefly deals with the situation in the USA since the end of 1988, describes interim solutions for current disposal requirements and then mainly concentrates on the German disposal concept currently being prepared. This concept initially envisages the long-term (25--50 years) dry interim storage of fuel elements in special containers in a central German interim store with subsequent direct final disposal without reprocessing of the irradiated fuel

  1. [Prevalence of dementia of insured persons with and without German citizenship : A study based on statuatory health insurance data].

    Science.gov (United States)

    Stock, Stephanie; Ihle, Peter; Simic, Dusan; Rupprecht, Christoph; Schubert, Ingrid; Lappe, Veronika; Kalbe, Elke; Tebest, Ralf; Lorrek, Kristina

    2018-04-01

    Elderly people with a non-German background are a fast growing population in Germany. Is administrative prevalence of dementia and uptake of nursing-home care similar in the German and non-German insured? Based on routine data, administrative prevalence rates for dementia were calculated for 2013 from a full census of data from one large sickness fund. Patients with dementia (PWD) were identified via ICD-10 codes (F00; F01; F03; F05; G30). Administrative prevalence of dementia was 2.67% in the study population; 3.06% in Germans, and 0.96% in non-Germans (p value German citizenship, except in women aged 80-84 (17.2 vs. 15.4) and for men in the age groups 80-84 (16.5 vs. 14.2), 85-89 years (23.4 vs. 21.5), and above 90 years of age (32.3 vs. 26.3). Standardized to the population of all investigated insured, 31.4% of all Germans with dementia had no longterm care entitlement vs. 35.5% of all patients without German citizenship. Of German patients, 55.1% were institutionalized vs. 39.5% of all patients without German citizenship. There was a higher prevalence of dementia in the very old insured without German citizenship compared to those with German citizenship, especially in men. Non-Germans showed lower uptake of nursing home care compared to Germans. Additionally, Germans had slightly higher nursing care entitlements. It should be investigated further how much of the difference is due to underdiagnosis, cultural differences, or lack of adequate diagnostic work-up.

  2. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

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

  3. Coursebook of German: Gender Aspect

    OpenAIRE

    Aleksandra Valeryevna Filippova

    2015-01-01

    The present article regards Aspekte 1 coursebook of German as a foreign language in the context of the gender policy initiated at the end of the last century by sociolinguists and by the representatives of the so called feminist criticism of the German language. This policy has been carried out up to date, and, according to many sociological and linguistic research, it is aimed at destructing gender stereotypes in teaching and reference materials. The use of this policy is conditioned by the ...

  4. Some Families of Asymmetric Quantum MDS Codes Constructed from Constacyclic Codes

    Science.gov (United States)

    Huang, Yuanyuan; Chen, Jianzhang; Feng, Chunhui; Chen, Riqing

    2018-02-01

    Quantum maximal-distance-separable (MDS) codes that satisfy quantum Singleton bound with different lengths have been constructed by some researchers. In this paper, seven families of asymmetric quantum MDS codes are constructed by using constacyclic codes. We weaken the case of Hermitian-dual containing codes that can be applied to construct asymmetric quantum MDS codes with parameters [[n,k,dz/dx

  5. Intelligibility of Standard German and Low German to Speakers of Dutch

    NARCIS (Netherlands)

    Gooskens, C.S.; Kürschner, Sebastian; van Bezooijen, R.

    2011-01-01

    This paper reports on the intelligibility of spoken Low German and Standard German for speakers of Dutch. Two aspects are considered. First, the relative potential for intelligibility of the Low German variety of Bremen and the High German variety of Modern Standard German for speakers of Dutch is

  6. Dictionary of high-energy physics English, German, French, Russian

    International Nuclear Information System (INIS)

    Sube, R.

    1987-01-01

    This volume contains nearly 4500 entries from branches of high-energy physics including cosmic radiation, elementary particles, elementary particle detection and measurement, field theories, and particle accelerators. Each English entry is numbered and followed by corresponding terms in the other languages. Alphabetical indexes of the German, French, and Russian terms are included

  7. 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)

  8. FAUST/CONTAIN

    International Nuclear Information System (INIS)

    Cherdron, W.; Minges, J.; Sauter, H.; Schuetz, W.

    1995-01-01

    The FAUNA facility has been restructured after completion of the sodium fire experiments. It is now serving LWR research, cf. report II on program no. 32.21.02 concerning steam explosions. The CONTAIN code system for computing the thermodynamic, aerosol and radiological phenomena in a containment under severe accident conditions is being developed with a new to fission product release and transport. (orig.)

  9. Containment vessel stability analysis

    International Nuclear Information System (INIS)

    Harstead, G.A.; Morris, N.F.; Unsal, A.I.

    1983-01-01

    The stability analysis for a steel containment shell is presented herein. The containment is a freestanding shell consisting of a vertical cylinder with a hemispherical dome. It is stiffened by large ring stiffeners and relatively small longitudinal stiffeners. The containment vessel is subjected to both static and dynamic loads which can cause buckling. These loads must be combined prior to their use in a stability analysis. The buckling loads were computed with the aid of the ASME Code case N-284 used in conjunction with general purpose computer codes and in-house programs. The equations contained in the Code case were used to compute the knockdown factors due to shell imperfections. After these knockdown factors were applied to the critical stress states determined by freezing the maximum dynamic stresses and combining them with other static stresses, a linear bifurcation analysis was carried out with the aid of the BOSOR4 program. Since the containment shell contained large penetrations, the Code case had to be supplemented by a local buckling analysis of the shell area surrounding the largest penetration. This analysis was carried out with the aid of the NASTRAN program. Although the factor of safety against buckling obtained in this analysis was satisfactory, it is claimed that the use of the Code case knockdown factors are unduly conservative when applied to the analysis of buckling around penetrations. (orig.)

  10. The general theory of convolutional codes

    Science.gov (United States)

    Mceliece, R. J.; Stanley, R. P.

    1993-01-01

    This article presents a self-contained introduction to the algebraic theory of convolutional codes. This introduction is partly a tutorial, but at the same time contains a number of new results which will prove useful for designers of advanced telecommunication systems. Among the new concepts introduced here are the Hilbert series for a convolutional code and the class of compact codes.

  11. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

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

  12. German radiological congress 1983

    International Nuclear Information System (INIS)

    Haubitz, B.; Stender, H.S.

    1983-01-01

    The publication contains the abstracts of the 261 papers read at the meeting and the 82 further papers announced, and 37 brief descriptions of the contributions to the scientific exhibition. The papers were on the subjects of radiology, nuclear medicine and to a certain extent, also radiobiology. (MG) [de

  13. New quantum codes constructed from quaternary BCH codes

    Science.gov (United States)

    Xu, Gen; Li, Ruihu; Guo, Luobin; Ma, Yuena

    2016-10-01

    In this paper, we firstly study construction of new quantum error-correcting codes (QECCs) from three classes of quaternary imprimitive BCH codes. As a result, the improved maximal designed distance of these narrow-sense imprimitive Hermitian dual-containing quaternary BCH codes are determined to be much larger than the result given according to Aly et al. (IEEE Trans Inf Theory 53:1183-1188, 2007) for each different code length. Thus, families of new QECCs are newly obtained, and the constructed QECCs have larger distance than those in the previous literature. Secondly, we apply a combinatorial construction to the imprimitive BCH codes with their corresponding primitive counterpart and construct many new linear quantum codes with good parameters, some of which have parameters exceeding the finite Gilbert-Varshamov bound for linear quantum codes.

  14. Quantum Codes From Cyclic Codes Over The Ring R 2

    International Nuclear Information System (INIS)

    Altinel, Alev; Güzeltepe, Murat

    2016-01-01

    Let R 2 denotes the ring F 2 + μF 2 + υ 2 + μυ F 2 + wF 2 + μwF 2 + υwF 2 + μυwF 2 . In this study, we construct quantum codes from cyclic codes over the ring R 2 , for arbitrary length n, with the restrictions μ 2 = 0, υ 2 = 0, w 2 = 0, μυ = υμ, μw = wμ, υw = wυ and μ (υw) = (μυ) w. Also, we give a necessary and sufficient condition for cyclic codes over R 2 that contains its dual. As a final point, we obtain the parameters of quantum error-correcting codes from cyclic codes over R 2 and we give an example of quantum error-correcting codes form cyclic codes over R 2 . (paper)

  15. Immobile Complex Verbs in Germanic

    DEFF Research Database (Denmark)

    Vikner, Sten

    2005-01-01

    the V° requirements or the V* requirements. Haider (1993, p. 62) and Koopman (1995), who also discuss such immobile verbs, only account for verbs with two prefix-like parts (e.g., German uraufführen ‘to perform (a play) for the first time' or Dutch herinvoeren ‘to reintroduce'), not for the more...... frequent type with only one prefix-like part (e.g., German bauchreden/Dutch buikspreken ‘to ventriloquize'). This analysis will try to account not only for the data discussed in Haider (1993) and Koopman (1995) but also for the following: - why immobile verbs include verbs with only one prefix-like part...... are immobile, - why such verbs are not found in Germanic VO-languages such as English and Scandinavian....

  16. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-08

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

  17. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

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

  18. Building codes : obstacle or opportunity?

    Science.gov (United States)

    Alberto Goetzl; David B. McKeever

    1999-01-01

    Building codes are critically important in the use of wood products for construction. The codes contain regulations that are prescriptive or performance related for various kinds of buildings and construction types. A prescriptive standard might dictate that a particular type of material be used in a given application. A performance standard requires that a particular...

  19. Radioactive action code

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    A new coding system, 'Hazrad', for buildings and transportation containers for alerting emergency services personnel to the presence of radioactive materials has been developed in the United Kingdom. The hazards of materials in the buildings or transport container, together with the recommended emergency action, are represented by a number of codes which are marked on the building or container and interpreted from a chart carried as a pocket-size guide. Buildings would be marked with the familiar yellow 'radioactive' trefoil, the written information 'Radioactive materials' and a list of isotopes. Under this the 'Hazrad' code would be written - three symbols to denote the relative radioactive risk (low, medium or high), the biological risk (also low, medium or high) and the third showing the type of radiation emitted, alpha, beta or gamma. The response cards indicate appropriate measures to take, eg for a high biological risk, Bio3, the wearing of a gas-tight protection suit is advised. The code and its uses are explained. (U.K.)

  20. PEAR code review

    International Nuclear Information System (INIS)

    De Wit, R.; Jamieson, T.; Lord, M.; Lafortune, J.F.

    1997-07-01

    As a necessary component in the continuous improvement and refinement of methodologies employed in the nuclear industry, regulatory agencies need to periodically evaluate these processes to improve confidence in results and ensure appropriate levels of safety are being achieved. The independent and objective review of industry-standard computer codes forms an essential part of this program. To this end, this work undertakes an in-depth review of the computer code PEAR (Public Exposures from Accidental Releases), developed by Atomic Energy of Canada Limited (AECL) to assess accidental releases from CANDU reactors. PEAR is based largely on the models contained in the Canadian Standards Association (CSA) N288.2-M91. This report presents the results of a detailed technical review of the PEAR code to identify any variations from the CSA standard and other supporting documentation, verify the source code, assess the quality of numerical models and results, and identify general strengths and weaknesses of the code. The version of the code employed in this review is the one which AECL intends to use for CANDU 9 safety analyses. (author)

  1. Capabilities of a mechanistic model for containment condenser simulation

    International Nuclear Information System (INIS)

    Broxtermann, Philipp; Cron, Daniel von der; Allelein, Hans-Josef

    2011-01-01

    In this paper the first application of the new containment COndenser MOdule (COMO) is being presented. COMO, just under development, represents a newly introduced part of the German containment code system COCOSYS and evaluates the contribution of containment condensers (CC) to passive containment cooling. Several next-generation Light Water Reactors (LWR) will feature innovative systems for passive heat removal during accidents from both the primary circuit and the containment. The passivity is based on natural driving forces only, such as gravity and natural circulation. To investigate the complex thermal-hydraulics and their propagation within a containment during accidents, containment code systems have been developed and validated against a wide variety of experiments. Furthermore, these codes are constantly improved to meet the intensified interest and knowledge in single phenomena and the technological evolution of the actual containment systems, e.g. the installation of passive devices. Accordingly, COCOSYS has been developed and is being continuously enhanced by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) and other partners like LRST. As has been shown by GRS and the calculation of the PANDA BC4 experiment, the interaction between a CC and the atmosphere of the surrounding vessel can be reproduced with the help of COCOSYS. However, up to now this has only been achieved on account of good knowledge of the outcome of the experiment, the user's skills and a complex input deck. The main goal of the newly introduced COMO is to improve the simulation of physical processes of CCs. This will be achieved by considering the passive driving forces which the CCs are based on. Up to now, a natural circulation within the CC tubes has been realized. The simulation of boiling conditions and the impact on the flow will be addressed in future studies. Additionally, the application of CCs to reactor simulation is being simplified and thus is supposed to reduce

  2. Aztheca Code

    International Nuclear Information System (INIS)

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

    2017-09-01

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

  3. Vocable Code

    DEFF Research Database (Denmark)

    Soon, Winnie; Cox, Geoff

    2018-01-01

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

  4. NSURE code

    International Nuclear Information System (INIS)

    Rattan, D.S.

    1993-11-01

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

  5. The mitochondrial genome of the German wasp Vespula germanica (Fabricius, 1793) (Hymenoptera: Vespoidea: Vespidae).

    Science.gov (United States)

    Zhou, Yuan; Hu, Yu-Lin; Xu, Zai-Fu; Wei, Shu-Jun

    2016-07-01

    The mitochondrial genome of the German wasp Vespula germanica (Fabricius, 1793) (Hymenoptera: Vespidae) (GenBank accession no. KR703583) was sequenced in the study. It represents the first mitochondrial genome from the genus Vespula. There are totally 163 42 bp in the currently sequenced portion of the genome, containing 13 protein-coding, two rRNA, and 18 tRNA genes and a partial A + T-rich region. Four tRNA genes of trnI, trnQ, trnM and trnY located at the downstream of the A + T-rich region were failed to sequence. At least two rearrangement events occurred in the sequenced region compared with the pupative ancestral arrangement of insects, corresponding to the translocation or remote inversion of tnnY from trnW-trnC-trnY cluster to the region of trnI-trnQ-trnM cluster and translocation of trnL1 from the downstream to the upstream of nad1 gene. All protein-coding genes start with ATN codons. Twelve and one protein-coding genes stop with termination codon TAA and T, respectively. Phylogenetic analysis using the Bayesian method based on all codon positions of the 13 protein-coding genes supports the monophyly of Vespidae and Formicidae. Within the Formicidae, the Myrmicinae and Formicinae form a sister group and then sister to the Dolichoderinae, while within the Vespidae, the Eumeninae sister to the lineage of Vespinae + Polistinae.

  6. The Danish Press during the German Occupation

    DEFF Research Database (Denmark)

    Roslyng-Jensen, Palle

    2010-01-01

    Censorship, self-censorship in Danish newspapers and Danish Radio during the German occupation of Denmark 1940-45......Censorship, self-censorship in Danish newspapers and Danish Radio during the German occupation of Denmark 1940-45...

  7. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

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

  8. Development of 3D models of buildings for containment of the nuclear power plant of Almaraz and of the Trillo Nuclear with the GOTHIC 8.0 code; Desarrollo de modelos 3D de los edificios de conten cion de la Central Nuclear de Almaraz y de la Central Nuclear de Trillo con el codigo GOTHIC 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Bocanegra Melian, R.; Fernandez Cosils, K.; Barreira Pereira, P.; Rey Peinado, L.; Posada Barral, J. M.

    2014-07-01

    The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)

  9. MELCOR computer code manuals

    Energy Technology Data Exchange (ETDEWEB)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.

  10. MELCOR computer code manuals

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

  11. A Suggested Curriculum Outline for German in Secondary Schools

    Science.gov (United States)

    Clutterbuck, J. M.

    1975-01-01

    Outlines a four-year program of German study aiming to give students a basic ability in spoken and written German, knowledge of German culture, and preparation for advanced German study. Study topics and textbooks are included. (CHK)

  12. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

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

  13. Uplink Coding

    Science.gov (United States)

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

    2007-01-01

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

  14. ANIMAL code

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1979-01-01

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

  15. Network Coding

    Indian Academy of Sciences (India)

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

  16. MCNP code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

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

  17. Expander Codes

    Indian Academy of Sciences (India)

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

  18. Metrical Phonology: German Sound System.

    Science.gov (United States)

    Tice, Bradley S.

    Metrical phonology, a linguistic process of phonological stress assessment and diagrammatic simplification of sentence and word stress, is discussed as it is found in the English and German languages. The objective is to promote use of metrical phonology as a tool for enhancing instruction in stress patterns in words and sentences, particularly in…

  19. Dividend Policy of German Firms

    NARCIS (Netherlands)

    Goergen, M.; Renneboog, L.D.R.; Correia Da Silva, L.

    2004-01-01

    German firms pay out a lower proportion of their cash flows than UK and US firms.However, on a published profits basis, the pattern is reversed.Company law provisions and accounting policies account for these conflicting results.A partial adjustment model is used to estimate the implicit target

  20. The German radiation protection standards

    International Nuclear Information System (INIS)

    Becker, Klaus; Neider, Rudolf

    1977-01-01

    The German Standards Institute (DIN Deutsches Institut fuer Normung, Berlin) is engaged in health physics standards development in the following committees. The Nuclear Standards Committee (NKe), which deals mainly with nuclear science and technology, the fuel cycle, and radiation protection techniques. The Radiology Standards Committee (FNR), whose responsibilities are traditionally the principles of radiation protection and dosimetry, applied medical dosimetry, and medical health physics. The German Electrotechnical Commission (DKE), which is concerned mostly with instrumentation standards. The Material Testing Committee (FNM), which is responsible for radiation protection in nonmedical radiography. The current body of over one hundred standards and draft standards was established to supplement the Federal German radiation protection legislation, because voluntary standards can deal in more detail with the specific practical problems. The number of standards is steadily expanding due to the vigorous efforts of about thirty working groups, consisting of essentially all leading German experts of this field. Work is supported by the industry and the Federal Government. A review of the present status and future plans, and of the international aspects with regard to European and world (ISO, etc.) standards will be presented

  1. Headstart German Program. Cultural Notes.

    Science.gov (United States)

    Defense Language Inst., Monterey, CA.

    This module provides cultural information that will be helpful to military personnel in understanding some aspects of the German way of life. The topics covered in the booklet are: housing, postal services, forms of address, courtesies, getting around, driving, hotels, restaurants, beer and wine, recreation, entertainment, health spas, shopping,…

  2. MARS Code in Linux Environment

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  3. MARS Code in Linux Environment

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Excessive Profits of German Defense Contractors

    Science.gov (United States)

    2014-09-01

    its business unit Thyssen Krupp Marine Systems, is a German defense contractor. (2) Tognom AG Tognum AG owned the MTU Friedrichshafen GmbH before... Friedrichshafen provided engines for many ships of the German Navy and for German battle tanks, such as the Leopard I and Leopard II. MTU refers to the

  5. [German influences on Romanian medical terminology].

    Science.gov (United States)

    Răcilă, R G; Răileanu, Irena; Rusu, V

    2008-01-01

    The medical terminology plays a key part both in the study of medicine as well as in its practice. Moreover, understanding the medical terms is important not only for the doctor but also for the patients who want to learn more about their condition. For these reasons we believe that the study of medical terminology is one of great interest. The aim of our paper was to evaluate the German linguistic and medical influences on the evolution of the Romanian medical terminology. Since the Romanian-German cultural contacts date back to the 12th century we had reasons to believe that the number of German medical words in Romanian would be significant. To our surprise, the Romanian language has very few German words and even less medical terms of German origin. However, when we searched the list of diseases coined after famous medical personalities, we found out that 26 % of them bore the names of German doctors and scientists. Taken together this proves that the German medical school played an important role on the evolution of Romanian medicine despite the fact that the Romanian vocabulary was slightly influenced by the German language. We explain this fact on the structural differences between the Romanian and German languages, which make it hard for German loans to be integrated in the Romanian lexis. In conclusion we state that the German influence on the Romanian medical terminology is weak despite the important contribution of the German medical school to the development of medical education and healthcare in Romania. Key

  6. Enriching the Curriculum with Pennsylvania German

    Science.gov (United States)

    Meindl, Joerg

    2016-01-01

    The German classroom should prepare students for the linguistic diversity of the target culture, including regional varieties and German spoken outside of the D-A-CH region. Because textbooks do not often include materials on regional varieties, this article presents a model to incorporate Pennsylvania German (PG) into the curriculum. The model…

  7. Silent Film in the German Classroom.

    Science.gov (United States)

    Caldwell, David

    In addition to using films in the German classroom to introduce students to German culture and history, it is important to show and study the film as film. This procedure emphasizes the importance of the film as a part of creative arts in Germany and demands student participation in observation and discussion. Many German silent films are…

  8. Stresses and strains in the steel containment resulting from transient pressure and temperature loading during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gruner, P.; Kuntze, W.M.; Jansky, J.

    1985-01-01

    Posttest calculations of stresses and strains in the steel containment of the German research reactor HDR were performed for a simulated LOCA. The results of the theoretical investigations are presented and compared to experimental findings. The pressure and temperature loading of the shell was determined with the thermodynamic code COFLOW on the basis of a multi-compartment model. Using a three-dimensional finite element model the temporal behaviour of the containment was calculated employing the structural mechanics code ASKA. Global bending deformations and local negative straining of the steel shell is discussed. Theoretical and experimental results agree in most cases rather well. Reasons for deviations will be discussed. The specific behaviour of strains found in the vicinity of locally heated areas will be explained by means of analytical considerations. (orig.)

  9. "We call it Springbok-German!": language contact in the German communities in South Africa.

    OpenAIRE

    Franke, Katharina

    2017-01-01

    Varieties of German are spoken all over the world, some of which have been maintained for prolonged periods of time. As a result, these transplanted varieties often show traces of the ongoing language contact as specific to their particular context. This thesis explores one such transplanted German language variety – Springbok- German – as spoken by a small subset of German Lutherans in South Africa. Specifically, this study takes as its focus eight rural German communities acr...

  10. Panda code

    International Nuclear Information System (INIS)

    Altomare, S.; Minton, G.

    1975-02-01

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

  11. Description of the COMRADEX code

    International Nuclear Information System (INIS)

    Spangler, G.W.; Boling, M.; Rhoades, W.A.; Willis, C.A.

    1967-01-01

    The COMRADEX Code is discussed briefly and instructions are provided for the use of the code. The subject code was developed for calculating doses from hypothetical power reactor accidents. It permits the user to analyze four successive levels of containment with time-varying leak rates. Filtration, cleanup, fallout and plateout in each containment shell can also be analyzed. The doses calculated include the direct gamma dose from the containment building, the internal doses to as many as 14 organs including the thyroid, bone, lung, etc. from inhaling the contaminated air, and the external gamma doses from the cloud. While further improvements are needed, such as a provision for calculating doses from fallout, rainout and washout, the present code capabilities have a wide range of applicability for reactor accident analysis

  12. CANAL code

    International Nuclear Information System (INIS)

    Gara, P.; Martin, E.

    1983-01-01

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

  13. The German Economy and U.S.-German Economic Relations

    Science.gov (United States)

    2009-11-30

    Should the SPD and The Left overcome existing differences, the grouping could represent a leftward shift in German politics. Alliance ’90 / The...and replaced it with less generous social assistance benefits already available to poor individuals, regardless of employment history . These changes...director at Volkswagen . 48Hans-Werner Sinn, Can Germany Be Saved?, p. 108. 49 Alister Miskimmon and Walter E. Paterson, “Conclusion: coping with the

  14. Hermitian self-dual quasi-abelian codes

    Directory of Open Access Journals (Sweden)

    Herbert S. Palines

    2017-12-01

    Full Text Available Quasi-abelian codes constitute an important class of linear codes containing theoretically and practically interesting codes such as quasi-cyclic codes, abelian codes, and cyclic codes. In particular, the sub-class consisting of 1-generator quasi-abelian codes contains large families of good codes. Based on the well-known decomposition of quasi-abelian codes, the characterization and enumeration of Hermitian self-dual quasi-abelian codes are given. In the case of 1-generator quasi-abelian codes, we offer necessary and sufficient conditions for such codes to be Hermitian self-dual and give a formula for the number of these codes. In the case where the underlying groups are some $p$-groups, the actual number of resulting Hermitian self-dual quasi-abelian codes are determined.

  15. Repertoires, Characters and Scenes: Sociolinguistic Difference in Turkish-German Comedy

    Science.gov (United States)

    Androutsopoulos, Jannis

    2012-01-01

    This paper examines representations of sociolinguistic difference in a German "ethnic comedy" as a means to contribute to a framework for the sociolinguistic study of film. Three levels of analysis of sociolinguistic difference in film are distinguished: repertoire analysis reconstructs the entirety of codes used in a film and their…

  16. Biotechnology 2000: a new German R&D programme

    OpenAIRE

    Ekkehard Warmuth

    1991-01-01

    Biotechnology 2000 is a German programme to continue the development of biotechnology started in 1982. It includes two new scientific fields for industrial innovation — genome research and neurobiology. Together with industry and the science community, the biotechnology programme will create a basis for future generations of biologically derived products and processes, including the development of safety precautions for the contained use of genetically modified organisms (GMOs) and of univers...

  17. 2014 German refrigeration and air conditioning meeting. Proceedings

    International Nuclear Information System (INIS)

    2014-01-01

    The proceedings of the 2014 German refrigeration and air conditioning meeting contain contributions on the following topics: cryotechnology, fundamentals and materials for the refrigeration and heat pump technology, devices and components for the refrigeration and heat pump technology, applications of refrigeration technologies, air conditioning technology and heat pump applications, cryotechnology in biology and medicine, heat transfer and ventilation, guidelines and legal topics, refrigerant fluid - oil mixtures, control and surveillance, simulation and control, ambient air.

  18. Validation of ASTEC core degradation and containment models

    International Nuclear Information System (INIS)

    Kruse, Philipp; Brähler, Thimo; Koch, Marco K.

    2014-01-01

    Ruhr-Universitaet Bochum performed in a German funded project validation of in-vessel and containment models of the integral code ASTEC V2, jointly developed by IRSN (France) and GRS (Germany). In this paper selected results of this validation are presented. In the in-vessel part, the main point of interest was the validation of the code capability concerning cladding oxidation and hydrogen generation. The ASTEC calculations of QUENCH experiments QUENCH-03 and QUENCH-11 show satisfactory results, despite of some necessary adjustments in the input deck. Furthermore, the oxidation models based on the Cathcart–Pawel and Urbanic–Heidrick correlations are not suitable for higher temperatures while the ASTEC model BEST-FIT based on the Prater–Courtright approach at high temperature gives reliable enough results. One part of the containment model validation was the assessment of three hydrogen combustion models of ASTEC against the experiment BMC Ix9. The simulation results of these models differ from each other and therefore the quality of the simulations depends on the characteristic of each model. Accordingly, the CPA FRONT model, corresponding to the simplest necessary input parameters, provides the best agreement to the experimental data

  19. Urban Green Infrastructure: German Experience

    OpenAIRE

    Diana Olegovna Dushkova; Sergey Nikolaevich Kirillov

    2016-01-01

    The paper presents a concept of urban green infrastructure and analyzes the features of its implementation in the urban development programmes of German cities. We analyzed the most shared articles devoted to the urban green infrastructure to see different approaches to definition of this term. It is based on materials of field research in the cities of Berlin and Leipzig in 2014-2015, international and national scientific publications. During the process of preparing the paper, consultations...

  20. German cross-cultural psychology

    OpenAIRE

    Trommsdorff, Gisela

    1986-01-01

    The present study deals with German-language cross-cultural research in different fields of psychology which attempts to achieve one Or more goals of cross-cultural psychology. First, methodological problems are discussed, followed by a selective presentation of cross-cultural research in personality, clinical, ethological, developmental, and social psychology. The theoretical and methodological advancement of these studies is investigated with respect to four approaches - universals in cross...

  1. German offshore wind turbine farms - status and prospective

    International Nuclear Information System (INIS)

    2004-08-01

    As a consequence of Germany's forthcoming phase-out of nuclear power the German government has initiated a number of activities in order to further development of renewable energy in the future. Offshore wind power has been chosen to play a central part. Although the first wind turbine has yet to be erected in German waters there is no doubt that it is a matter of time before the growing German market will gather speed. The objective of this report is to provide Danish business enterprises with interests in wind power with an insight into the German offshore wind power market and the export possibilities of the present and in the near future. As introduction the report lists the general outlines for construction and operation of wind turbine farms in Germany, furthermore, a number of additional conditions that Danish business enterprises should be aware of are listed. The introduction is followed by an up-to-the -minute status account of all ongoing projects. This part of the report has been made on the basis of a questionnaire send out by the Danish Embassy to project leaders in the business enterprises behind the project planning. Finally, the report provides an overview of all partners behind the planned wind farms. The overview contains contact information as well as information about the composition of project companies and consortiums. (BA)

  2. Computer codes used in particle accelerator design: First edition

    International Nuclear Information System (INIS)

    1987-01-01

    This paper contains a listing of more than 150 programs that have been used in the design and analysis of accelerators. Given on each citation are person to contact, classification of the computer code, publications describing the code, computer and language runned on, and a short description of the code. Codes are indexed by subject, person to contact, and code acronym

  3. The Role of Structural Funding for Stability in the German Banking Sector

    OpenAIRE

    Schupp, Fabian; Silbermann, Leonid

    2017-01-01

    We analyze whether, and if so by how much, stable funding would have contributed to the financial soundness of German banks in the time period between 1995 and 2013, before the Basel III liquidity regulation to address excessive maturity mismatches in the wake of the financial crisis via the Net Stable Funding Ratio can be expected to have been fully implemented. Using a dataset that contains information on critical events of German banks, we find that financing loans using fewer customer dep...

  4. New quantum codes derived from a family of antiprimitive BCH codes

    Science.gov (United States)

    Liu, Yang; Li, Ruihu; Lü, Liangdong; Guo, Luobin

    The Bose-Chaudhuri-Hocquenghem (BCH) codes have been studied for more than 57 years and have found wide application in classical communication system and quantum information theory. In this paper, we study the construction of quantum codes from a family of q2-ary BCH codes with length n=q2m+1 (also called antiprimitive BCH codes in the literature), where q≥4 is a power of 2 and m≥2. By a detailed analysis of some useful properties about q2-ary cyclotomic cosets modulo n, Hermitian dual-containing conditions for a family of non-narrow-sense antiprimitive BCH codes are presented, which are similar to those of q2-ary primitive BCH codes. Consequently, via Hermitian Construction, a family of new quantum codes can be derived from these dual-containing BCH codes. Some of these new antiprimitive quantum BCH codes are comparable with those derived from primitive BCH codes.

  5. From concatenated codes to graph codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2004-01-01

    We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...

  6. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

    This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

  7. Effective enforcement of the forest practices code

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    The British Columbia Forest Practices Code establishes a scheme to guide and direct forest harvesting and other forest uses in concert with other related acts. The Code is made up of the Forest Practices Code of British Columbia Act, regulations, standards, and guidebooks. This document provides information on Code enforcement. It reviews the roles of the three provincial resource ministries and the Attorney General in enforcing the code, the various activities undertaken to ensure compliance (including inspections, investigations, and responses to noncompliance), and the role of the public in helping to enforce the Code. The appendix contains a list of Ministry of Forests office locations and telephone numbers.

  8. Standardized Definitions for Code Verification Test Problems

    Energy Technology Data Exchange (ETDEWEB)

    Doebling, Scott William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-14

    This document contains standardized definitions for several commonly used code verification test problems. These definitions are intended to contain sufficient information to set up the test problem in a computational physics code. These definitions are intended to be used in conjunction with exact solutions to these problems generated using Exact- Pack, www.github.com/lanl/exactpack.

  9. The intercomparison of aerosol codes

    International Nuclear Information System (INIS)

    Dunbar, I.H.; Fermandjian, J.; Gauvain, J.

    1988-01-01

    The behavior of aerosols in a reactor containment vessel following a severe accident could be an important determinant of the accident source term to the environment. Various processes result in the deposition of the aerosol onto surfaces within the containment, from where they are much less likely to be released. Some of these processes are very sensitive to particle size, so it is important to model the aerosol growth processes: agglomeration and condensation. A number of computer codes have been written to model growth and deposition processes. They have been tested against each other in a series of code comparison exercises. These exercises have investigated sensitivities to physical and numerical assumptions and have also proved a useful means of quality control for the codes. Various exercises in which code predictions are compared with experimental results are now under way

  10. What Froze the Genetic Code?

    Directory of Open Access Journals (Sweden)

    Lluís Ribas de Pouplana

    2017-04-01

    Full Text Available The frozen accident theory of the Genetic Code was a proposal by Francis Crick that attempted to explain the universal nature of the Genetic Code and the fact that it only contains information for twenty amino acids. Fifty years later, it is clear that variations to the universal Genetic Code exist in nature and that translation is not limited to twenty amino acids. However, given the astonishing diversity of life on earth, and the extended evolutionary time that has taken place since the emergence of the extant Genetic Code, the idea that the translation apparatus is for the most part immobile remains true. Here, we will offer a potential explanation to the reason why the code has remained mostly stable for over three billion years, and discuss some of the mechanisms that allow species to overcome the intrinsic functional limitations of the protein synthesis machinery.

  11. What Froze the Genetic Code?

    Science.gov (United States)

    Ribas de Pouplana, Lluís; Torres, Adrian Gabriel; Rafels-Ybern, Àlbert

    2017-04-05

    The frozen accident theory of the Genetic Code was a proposal by Francis Crick that attempted to explain the universal nature of the Genetic Code and the fact that it only contains information for twenty amino acids. Fifty years later, it is clear that variations to the universal Genetic Code exist in nature and that translation is not limited to twenty amino acids. However, given the astonishing diversity of life on earth, and the extended evolutionary time that has taken place since the emergence of the extant Genetic Code, the idea that the translation apparatus is for the most part immobile remains true. Here, we will offer a potential explanation to the reason why the code has remained mostly stable for over three billion years, and discuss some of the mechanisms that allow species to overcome the intrinsic functional limitations of the protein synthesis machinery.

  12. Model of the containment building of Almaraz NPP and the system of recombiners PARs, with the GOTHIC code, for the study of the diffusion of combustible gases; Modelo del edificio de contencion de C.N. Almaraz y del sistema de recombinadores PARs, con el codigo GOTHIC, para el estudio de la difusion de gases combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Gonzalez, M.; Huelamo, E.; Mazrtinez, M.; Perez, J. R.

    2014-07-01

    This paper presents the analysis of distribution of gases within the containment building carried out a simulation model with the code Thermo hydraulic GOTHIC, which has been evaluated based on passive autocatalytic recombiners gas control system. The model considers scenarios of severe accident with specific conditions that produce the most hydrogen generation rates. Intended to verify the effectiveness of the control system of gas expected to be installed in the Almaraz Nuclear power plant so that the number and location of recombiners equipment meets its function of preventing the formation of explosive atmospheres which impairs the integrity of the containment, reducing and limiting the concentration of combustible gases during the postulated accident. (Author)

  13. Operational Art and the German 1918 Offensives

    OpenAIRE

    Zabecki, D T

    2009-01-01

    At the tactical level of war the Germans are widely regarded as having had the most innovative and proficient army of World War I. Likewise, many historians would agree that the Germans suffered from serious, if not fatal, shortcomings at the strategic level of war. It is at the middle level of warfare, the operational level, that the Germans seem to be the most difficult to evaluate. Although the operational was only fully accepted in the 1980s by many Western militaries as...

  14. Out of the German parliament into the German Museum?

    International Nuclear Information System (INIS)

    Lieb, E.

    1989-01-01

    It is currently discussed whether the German Bundestag can deal with the interdepartmental problems of technology assessment with the Commissions of Inquiry on the one hand and whether it has adequate instruments available with the department-related standing Bundestag committees in order to deal with technology assessment. In its report the Commission of Inquiry for Technology Assessment of the past legislative period came to the conclusion that the US parliamentary advisory model which has been realized with OTA could, of course, not be transferred to the situation of the German Bundestag without hesitation, but that the Bundestag should also have a permanent scientific department staff with a sufficient number of personnel and material. The congress was to offer the possibility to discuss the problems of technology assessment with regard to this up-to-date background with experts and members of parliament of the various commissions of inquiry and commissions of the Bundestag which were summoned in order to judge essential technologies and also to solve the problem of the institutionalization of technology assessment. (orig./DG) [de

  15. Coursebook of German: Gender Aspect

    Directory of Open Access Journals (Sweden)

    Aleksandra Valeryevna Filippova

    2015-09-01

    Full Text Available The present article regards Aspekte 1 coursebook of German as a foreign language in the context of the gender policy initiated at the end of the last century by sociolinguists and by the representatives of the so called feminist criticism of the German language. This policy has been carried out up to date, and, according to many sociological and linguistic research, it is aimed at destructing gender stereotypes in teaching and reference materials. The use of this policy is conditioned by the fact that there is a problem of women discrimination in the textbooks, which provide classical gender stereotypes, where, in spite of modern social changes, women are still overrepresented in the private domain and underrepresented in the public sphere. Apart from that, gender stereotypes and gender asymmetry are embedded in the language, where the woman is often not referred to directly while the man is used in the generalizing meaning of "human". The gender asymmetry is reflected in the idioms as well. Nevertheless the analysis of modern coursebooks reveals both some changes in the presentation of women and men's occupations and in the language due to the usage of so-called "gender neutral" forms. The objective of our research lies in the linguistic analysis of the usage of the "gender neutral" forms as well as in the coursebooks on phraseology in order to find out gender asymmetries. In addition, the author focuses on gender stereotypes, men and women's behavioral patterns in different domains of life, and positive changes in the image of men and women represented in the Aspekte German coursebook.

  16. German Policy Towards Muslim Communities

    Directory of Open Access Journals (Sweden)

    Liudmila R. Sadykova

    2014-01-01

    Full Text Available The past two-three decades can be characterized by the period of global migration and sharp jump of migratory streams is connected with globalization and with the economic factor, generating labor movement behind resources from Third World countries to the countries with deficiency of labor. The desire to receive comfort life becomes the major reason, and the migrant makes the decision being guided by private interest more often instead of external factors. Western Europe became one of the most important center of gravity of migrants. During the post-war period the need of Europe in foreign labor for restoration of the economy destroyed by war, laid the foundation of mass international migration to this region. Globalization of migratory streams, penetration of foreign culture groups into structure of accepting society and prevalence of multicultural, multiethnic societies are important characteristics of a modern era. Western Europe became one of the most important centers of gravity of migrants. During the post-war period, the need of Europe in foreign labor for restoration of the economy destroyed by war laid the foundation of mass international migration to this region. Special relevance the problem of reception of immigrants, in particular from the Muslim countries, got for the former colonial powers, in particular Great Britain, France, and the Netherlands. Germany also faced this problem; migrants workers from other countries were required for the post-war restoration. Now Germany still is one of the main centers of an attraction of migrants, and concentration of them in this country annually increases. Despite the steps taken by the German government on elimination of Muslim isolation in the German society, its efforts did not bear fruits so far. The majority of Muslims live their life and are still torn off from high life of the country. A possible threat of destruction of the German community appeared when the various ethnic groups

  17. Pressurized thermal shock analysis in German nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Fricke, Stefan; Braun, Michael [TUEV NORD Nuclear, Hannover (Germany)

    2015-03-15

    For more than 30 years TUeV NORD is a competent consultant in nuclear safety is-sues giving expert third party opinion to our clients. According to the German regulations the safety against brittle fracture has to be proved for the reactor pressure vessel (RPV) and with a new level of knowledge the proof has to be continuously updated with the development in international codes and standards like ASME, BS and RCC-M. The load of the RPV is a very complex transient pressure and temperature situation. Today these loading conditions can be modeled by thermal hydraulic calculations and new experimental results much more detailed than in the construction phase of German Nuclear Power Plants in the 1980s. Therefore, the proof against brittle fracture from the construction phase had to be updated for all German Nuclear Power Plants with the new findings of the loading conditions especially for a postulated small leakage in the main coolant line. The RPV consists of ferritic base material (about 250 mm) and austenitic cladding (about 6 mm) at the inner side. The base material and the cladding have different physical properties which have to be considered temperature dependently in the cal-culations. Radiation-embrittlement effects on the material are to be respected in the fracture mechanics assessment. The regions of the RPV of special interest are the core weld, the inlet and outlet nozzle region and the flange connecting weld zone. The fracture mechanics assessment is performed for normal and abnormal operating conditions and for accidents like LOCA (Loss of Coolant Accident). In this paper the German approach to fracture mechanics assessment to brittle fracture will be discussed from the point of view of a third party organization.

  18. Genetic drift. Overview of German, Nazi, and Holocaust medicine.

    Science.gov (United States)

    Cohen, M Michael

    2010-03-01

    An overview of German, Nazi, and Holocaust medicine brings together a group of subjects discussed separately elsewhere. Topics considered include German medicine before and during the Nazi era, such as advanced concepts in epidemiology, preventive medicine, public health policy, screening programs, occupational health laws, compensation for certain medical conditions, and two remarkable guidelines for informed consent for medical procedures; also considered are the Nuremberg Code; American models for early Nazi programs, including compulsory sterilization, abusive medical experiments on prison inmates, and discrimination against black people; two ironies in US and Nazi laws; social Darwinism and racial hygiene; complicity of Nazi physicians, including the acts of sterilization, human experimentation, and genocide; Nazi persecution of Jewish physicians; eponyms of unethical German physicians with particular emphasis on Reiter, Hallervorden, and Pernkopf; eponyms of famous physicians who were Nazi victims, including Pick and van Creveld; and finally, a recommendation for convening an international committee of physicians and ethicists to deal with five issues: (a) to propose alternative names for eponyms of physicians who exhibited complicity during the Nazi era; (b) to honor the eponyms and stories of physicians who were victims of Nazi atrocities and genocide; (c) to apply vigorous pressure to those German and Austrian Institutes that have not yet undertaken investigations to determine if the bodies of Nazi victims remain in their collections; (d) to recommend holding annual commemorations in medical schools and research institutes worldwide to remember and to reflect on the victims of compromised medical practice, particularly, but not exclusively, during the Nazi era because atrocities and acts of genocide have occurred elsewhere; and (e) to examine the influence of any political ideology that compromises the practice of medicine. (c) 2010 Wiley-Liss, Inc

  19. German risk study of PWR's

    International Nuclear Information System (INIS)

    Kafka, P.

    1983-01-01

    In this paper, first the status of German Risk Study is presented briefly. Specific reference is made to the investigations in Phase B of the study and related programs. Significant elements involved in the risk assessment for NPPs, mainly in the field of system and structural reliability analyses are mentioned. In particular, important outcomes and limiting facts in the process of a Probabilistic Risk Assessment (PRA) to evaluate the safety standard and above all the influence of individual components or subsystems on core melt frequency are discussed. (orig.)

  20. German Librarianship and Munich Libraries

    Directory of Open Access Journals (Sweden)

    Osman Ümit Özen

    1994-06-01

    Full Text Available There are 27 municipal libraries including the Central Public Library in Munich. The other important libraries in the city are Bayern State National Library, Maximillian University Library, a technical highschool library and the "Deutsches Musuem" Library. All these libraries are financed locally. The author introduces these libraries briefly and compares German libraries with Turkish libraries. He concludes that although theoretically there are not distinctive differences, in practice, buildings and their layout are better in Germany where more variety of services are offered. In Turkey standardization has not been realized yet. Turkey needs to computerize and network to improve the services offered in an efficient way.

  1. Iterative nonlinear unfolding code: TWOGO

    International Nuclear Information System (INIS)

    Hajnal, F.

    1981-03-01

    a new iterative unfolding code, TWOGO, was developed to analyze Bonner sphere neutron measurements. The code includes two different unfolding schemes which alternate on successive iterations. The iterative process can be terminated either when the ratio of the coefficient of variations in terms of the measured and calculated responses is unity, or when the percentage difference between the measured and evaluated sphere responses is less than the average measurement error. The code was extensively tested with various known spectra and real multisphere neutron measurements which were performed inside the containments of pressurized water reactors

  2. [Orthopedic and trauma surgery in the German-DRG-System 2009].

    Science.gov (United States)

    Franz, D; Windolf, J; Siebert, C H; Roeder, N

    2009-01-01

    The German DRG-System was advanced into version 2009. For orthopedic and trauma surgery significant changes concerning coding of diagnoses, medical procedures and concerning the DRG-structure were made. Analysis of relevant diagnoses, medical procedures and G-DRGs in the versions 2008 and 2009 based on the publications of the German DRG-institute (InEK) and the German Institute of Medical Documentation and Information (DIMDI). Changes for 2009 focussed on the development of DRG-structure, DRG-validation and codes for medical procedures to be used for very complex cases. The outcome of these changes for German hospitals may vary depending in the range of activities. G-DRG-System gained complexity again. High demands are made on correct and complete coding of complex orthopedic and trauma surgery cases. Quality of case-allocation within the G-DRG-System was improved. Nevertheless, further adjustments of the G-DRG-System especially for cases with severe injuries are necessary.

  3. The German social democratic party (SPD) and the debate on the fertility decline in the German Empire (1870~1918).

    Science.gov (United States)

    Mun, Soo-Hyun

    2011-12-31

    servitude for anyone who had an abortion or people who helped to practice it, Paragraph 184.3 of the civil code was enacted in order to outlaw the advertising, display, and publicizing of contraceptives with an 'indecent' intention, although selling or manufacturing contraceptives was not forbidden. Such a punitive approach was especially preferred by the government and conservative parties because it was easy to implement and "cheap" in comparison with the comprehensive social welfare program. What made the SPD different from other conservative parties was the fact that the SPD opposed the government's attempt to prohibit contraception by means of strengthening a penal code. According to the SPD, it was not only morally unacceptable, but also technically impossible for the government to intervene in family limitation. Moreover, politicians from the SPD criticized that such a punitive policy targeted the working class because the upper echelon of the society could easily evade the ban on contraceptives. However, the SPD did not proceed to draft comprehensive social welfare measures in order to fight the fertility decline. The miserable condition of working class women remained as an invisible social phenomenon even within the SPD. The German women who could not find the proper means to practice contraception were driven to have abortions. Annually, hundreds of the women were accused of practicing abortion and imprisoned. In sum, German society ran about in confusion and did not know how to properly respond to the unprecedented decline in fertility. By defining the fertility decline just as a social disease due to moral decay and influence of socialism, German society lost a chance to rationalize itself. Given that women, the main actors, had no way to take part in the debate over this issue, it is not surprising that German society fought against the symptom of the disease, not against its root.

  4. Towers of generalized divisible quantum codes

    Science.gov (United States)

    Haah, Jeongwan

    2018-04-01

    A divisible binary classical code is one in which every code word has weight divisible by a fixed integer. If the divisor is 2ν for a positive integer ν , then one can construct a Calderbank-Shor-Steane (CSS) code, where X -stabilizer space is the divisible classical code, that admits a transversal gate in the ν th level of Clifford hierarchy. We consider a generalization of the divisibility by allowing a coefficient vector of odd integers with which every code word has zero dot product modulo the divisor. In this generalized sense, we construct a CSS code with divisor 2ν +1 and code distance d from any CSS code of code distance d and divisor 2ν where the transversal X is a nontrivial logical operator. The encoding rate of the new code is approximately d times smaller than that of the old code. In particular, for large d and ν ≥2 , our construction yields a CSS code of parameters [[O (dν -1) ,Ω (d ) ,d ] ] admitting a transversal gate at the ν th level of Clifford hierarchy. For our construction we introduce a conversion from magic state distillation protocols based on Clifford measurements to those based on codes with transversal T gates. Our tower contains, as a subclass, generalized triply even CSS codes that have appeared in so-called gauge fixing or code switching methods.

  5. A German catastrophe? German historians and the Allied bombings, 1945-2010

    NARCIS (Netherlands)

    von Benda-Beckmann, B.R.

    2010-01-01

    As one of the major symbols of German suffering, the Allied bombing war left a strong imprint on German society. To a much wider extent than is often claimed, the Allied bombings became part of German debates on the Second World War. In both the GDR as well as the Federal Republic before and after

  6. The French-German common safety approach for future reactors

    International Nuclear Information System (INIS)

    Birkhofer, A.; Chevet, P.F.

    1995-01-01

    A common safety approach has been defined for future electronuclear plants in the framework of the French-German European Pressurised water Reactor (EPR) project. Improvements in the domain of containment are required in future reactors conception to prevent any risk of core fusion under high and low pressure. Another objective is to reduce significantly the radioactive releases due to other accidents in order to reduce spatial and temporal environmental and human protection procedures. Protection against external aggressions (plane fall, explosions, earthquakes,..), prevention of pipe rupture in the primary circuits, limitation of hydrogen production in the case of water-zirconium complete reaction, cooling of the reactor in the case of core fusion, and radiologic consequences of accidents are the main points discussed by the French-German safety authorities to define the common safety standards of the EPR project. (J.S.)

  7. The Atomic Law, the German Bundesrat and the administrative organisation

    International Nuclear Information System (INIS)

    Burgi, Martin

    2011-01-01

    Soon, the Federal Constitutional Court (Karlsruhe, Federal Republic of Germany) will deal with both the Eleventh Amendment of the Atomic Energy Act effecting the extension of the operating period of nuclear power plants as well as with the Twelfth Amendment of the Atomic Energy Act which in particular contains some security-related regulations due to European legal occasion. The emphasis is on the Article 87c of the Basic Law. According to Article 87c of the Basic Law, the legislation in the field of nuclear law requires the consent of the German Bundesrat. The possible of approval of both laws is subject to certain administrative organization legal circumstances. The sober investigation and evaluation of these circumstances in the context of Article 83 et seq. of the Basic Law results to the conclusion that the two amending laws do not require the consent of the German Bundesrat.

  8. Consequence model of the German reactor safety study

    International Nuclear Information System (INIS)

    Bayer, A.; Aldrich, D.; Burkart, K.; Horsch, F.; Hubschmann, W.; Schueckler, M.; Vogt, S.

    1979-01-01

    The consequency model developed for phase A of the German Reactor Safety Study (RSS) is similar in many respects to its counterpart in WASH-1400. As in that previous study, the model describes the atmosphere dispersion and transport of radioactive material released from the containment during a postulated reactor accident, and predicts its interaction with and influence on man. Differences do exist between the two models however, for the following reasons: (1) to more adequately reflect central European conditions, (2) to include improved submodels, and (3) to apply additional data and knowledge that have become available since publication of WASH-1400. The consequence model as used in phase A of the German RSS is described, highlighting differences between it and the U.S. model

  9. Automatic coding method of the ACR Code

    International Nuclear Information System (INIS)

    Park, Kwi Ae; Ihm, Jong Sool; Ahn, Woo Hyun; Baik, Seung Kook; Choi, Han Yong; Kim, Bong Gi

    1993-01-01

    The authors developed a computer program for automatic coding of ACR(American College of Radiology) code. The automatic coding of the ACR code is essential for computerization of the data in the department of radiology. This program was written in foxbase language and has been used for automatic coding of diagnosis in the Department of Radiology, Wallace Memorial Baptist since May 1992. The ACR dictionary files consisted of 11 files, one for the organ code and the others for the pathology code. The organ code was obtained by typing organ name or code number itself among the upper and lower level codes of the selected one that were simultaneous displayed on the screen. According to the first number of the selected organ code, the corresponding pathology code file was chosen automatically. By the similar fashion of organ code selection, the proper pathologic dode was obtained. An example of obtained ACR code is '131.3661'. This procedure was reproducible regardless of the number of fields of data. Because this program was written in 'User's Defined Function' from, decoding of the stored ACR code was achieved by this same program and incorporation of this program into program in to another data processing was possible. This program had merits of simple operation, accurate and detail coding, and easy adjustment for another program. Therefore, this program can be used for automation of routine work in the department of radiology

  10. French pollution and German lignite

    International Nuclear Information System (INIS)

    Foos, Jacques

    2015-01-01

    After having recalled that the German energy transition is based on a complete shutting down of nuclear power stations to replace them by renewable energy sources on the one hand, and by coal (lignite, i.e. the dirtiest coal) and gas on the other hand to compensate the intermittency of the former ones, this article notices that pollution peaks occurred in France when an eastern of north-eastern wind was blowing, and not in case of western wind. The author then wanders whether this pollution comes from Germany, and more particularly from the releases of lignite-fuelled power stations. Then, the author comments the high level of pollution associated with coal extraction and exploitation in Germany, causing thousands of deaths and resulting in lung diseases or cancers, myocardial infractions. The author then makes a parallel between, on the one hand, the ignorance of this German pollution and, on the other hand, evacuation measures around Fukushima for a radioactivity which the author considers as less dangerous in terms of life expectancy

  11. Error-correction coding

    Science.gov (United States)

    Hinds, Erold W. (Principal Investigator)

    1996-01-01

    This report describes the progress made towards the completion of a specific task on error-correcting coding. The proposed research consisted of investigating the use of modulation block codes as the inner code of a concatenated coding system in order to improve the overall space link communications performance. The study proposed to identify and analyze candidate codes that will complement the performance of the overall coding system which uses the interleaved RS (255,223) code as the outer code.

  12. Reducing barriers to energy efficiency in the German higher education sector. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Schleich, J.; Boede, U.

    2000-12-01

    This report describes the empirical research into barriers to energy efficiency in the German higher education (HE) sector. It is one of nine such reports in the BARRIERS project. The report contains description and analysis of six case studies of energy management in German universities. The results are analysed using the theoretical framework developed for the BARRIERS project (Sorrell et al., 2000). The report also provides brief recommendations on how these barriers to the rational use of energy (RUE) may be overcome and how energy efficiency within the sector may be improved. The results of the study for the higher education sector in Germany are summarised in this executive summary under the following headings: - Characterising the higher education sector; - Case studies of energy management in the German higher education sector; - Evidence of barriers in the German higher education sector; - The role of energy service companies in the higher education sector; - Policy implications. (orig.)

  13. Reducing barriers to energy efficiency in the German higher education sector. Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Schleich, J.; Boede, U.

    2000-12-01

    This report describes the empirical research into barriers to energy efficiency in the German higher education (HE) sector. It is one of nine such reports in the BARRIERS project. The report contains description and analysis of six case studies of energy management in German universities. The results are analysed using the theoretical framework developed for the BARRIERS project (Sorrell et al., 2000). The report also provides brief recommendations on how these barriers to the rational use of energy (RUE) may be overcome and how energy efficiency within the sector may be improved. The results of the study for the higher education sector in Germany are summarised in this executive summary under the following headings: - Characterising the higher education sector; - Case studies of energy management in the German higher education sector; - Evidence of barriers in the German higher education sector; - The role of energy service companies in the higher education sector; - Policy implications. (orig.)

  14. Reducing barriers to energy efficiency in the German mechanical engineering sector. Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Schleich, J.; Boede, U.

    2000-12-01

    This report describes the empirical research into barriers to energy efficiency in the German mechanical engineering (ME) sector. It is one of nine such reports in the BARRIERS project. The report contains description and analysis of four case studies of energy management in German companies in the ME sector. The results are analysed using the theoretical framework developed for the BARRIERS project. The report also provides brief recommendations on how these barriers to the rational use of energy (RUE) may be overcome and how energy efficiency within the ME sector may be improved. The results of the study for the ME sector in Germany are summarised in this executive summary under the following headings: - Characterising the mechanical engineering sector; - Case studies of energy management in the German mechanical engineering sector; - Evidence of barriers in the German mechanical engineering sector; - The role of energy service companies in the mechanical engineering sector; - Policy implications. (orig.)

  15. Reducing barriers to energy efficiency in the German brewing sector. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Schleich, J; Boede, U; Ostertag, K; Radgen, P

    2000-12-01

    This report describes the empirical research into barriers to energy efficiency in the German brewing sector. It is one of nine such reports in the BARRIERS project. The report contains description and analysis of five case studies of energy management in German breweries. The results are analysed using the theoretical framework developed for the BARRIERS project. The report also provides brief recommendations on how these barriers to the rational use of energy (RUE) may be overcome and how energy efficiency within the brewing sector may be improved. The results of the study for the brewing sector in Germany are summarised in this executive summary under the following headings: - Characterising the brewing sector - Case studies of energy management in the German brewing sector; - Evidence of barriers in the German brewing sector; - The role of energy service companies in the brewing sector; - Policy implications. (orig.)

  16. Reducing barriers to energy efficiency in the German mechanical engineering sector. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Schleich, J.; Boede, U.

    2000-12-01

    This report describes the empirical research into barriers to energy efficiency in the German mechanical engineering (ME) sector. It is one of nine such reports in the BARRIERS project. The report contains description and analysis of four case studies of energy management in German companies in the ME sector. The results are analysed using the theoretical framework developed for the BARRIERS project. The report also provides brief recommendations on how these barriers to the rational use of energy (RUE) may be overcome and how energy efficiency within the ME sector may be improved. The results of the study for the ME sector in Germany are summarised in this executive summary under the following headings: - Characterising the mechanical engineering sector; - Case studies of energy management in the German mechanical engineering sector; - Evidence of barriers in the German mechanical engineering sector; - The role of energy service companies in the mechanical engineering sector; - Policy implications. (orig.)

  17. Reducing barriers to energy efficiency in the German brewing sector. Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Schleich, J.; Boede, U.; Ostertag, K.; Radgen, P.

    2000-12-01

    This report describes the empirical research into barriers to energy efficiency in the German brewing sector. It is one of nine such reports in the BARRIERS project. The report contains description and analysis of five case studies of energy management in German breweries. The results are analysed using the theoretical framework developed for the BARRIERS project. The report also provides brief recommendations on how these barriers to the rational use of energy (RUE) may be overcome and how energy efficiency within the brewing sector may be improved. The results of the study for the brewing sector in Germany are summarised in this executive summary under the following headings: - Characterising the brewing sector; - Case studies of energy management in the German brewing sector; - Evidence of barriers in the German brewing sector; - The role of energy service companies in the brewing sector; - Policy implications. (orig.)

  18. LFSC - Linac Feedback Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Valentin; /Fermilab

    2008-05-01

    The computer program LFSC (Code>) is a numerical tool for simulation beam based feedback in high performance linacs. The code LFSC is based on the earlier version developed by a collective of authors at SLAC (L.Hendrickson, R. McEwen, T. Himel, H. Shoaee, S. Shah, P. Emma, P. Schultz) during 1990-2005. That code was successively used in simulation of SLC, TESLA, CLIC and NLC projects. It can simulate as pulse-to-pulse feedback on timescale corresponding to 5-100 Hz, as slower feedbacks, operating in the 0.1-1 Hz range in the Main Linac and Beam Delivery System. The code LFSC is running under Matlab for MS Windows operating system. It contains about 30,000 lines of source code in more than 260 subroutines. The code uses the LIAR ('Linear Accelerator Research code') for particle tracking under ground motion and technical noise perturbations. It uses the Guinea Pig code to simulate the luminosity performance. A set of input files includes the lattice description (XSIF format), and plane text files with numerical parameters, wake fields, ground motion data etc. The Matlab environment provides a flexible system for graphical output.

  19. Improved hydrogen distribution calculation in the containment using the coupled codes MELCOR and GASFLOW for the analysis of severe accidents in nuclear power plants; Verbesserte Berechnung der Wasserstoffverteilung im Sicherheitsbehaelter bei der Analyse schwerer Stoerfaelle in Kernkraftwerken durch Kopplung von MELCOR und GASFLOW

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, Tobias

    2014-09-01

    The risk of a hydrogen combustion within a containment of a pressurized water reactor during a severe loss of coolant accident (LOCA) is evaluated using numerical simulations. The code MELCOR provides integral analysis capabilities for severe accidents. Yet, its Lumped Parameter (LP) model provides less accurate information about on thermal hydraulics within the containment during a LOCA. GASFLOW is a CFD code that simulates both the local hydrogen distribution and the pressure inside the containment more realistically. Currently, to perform these GASFLOW simulations, the common procedure is to use a source term from a previous MELCOR calculation. However, with this approach, the influence of the more realistic GASFLOW pressure on the mass flow through the leak cannot be taken into account. This inconsistency is overcome by coupling both codes in this thesis. Here, the MELCOR instance is responsible for the primary and secondary systems. At the same time, the GASFLOW instance predicts the thermal hydraulics of the containment. The more accurate containment pressure from the GASFLOW instance is used in the MELCOR instance to calculate consistent outflow rates through the leak. In order to couple both codes, the existing interface in MELCOR is modified and a new interface for GASFLOW is developed and implemented. To begin with, the hydrogen distribution inside a generic containment is calculated by MELCOR using a typical coarse LP nodalization and a refined one. The results obtained are compared to a GASFLOW simulation. It is shown that the refinement only leads to a better agreement with the GASFLOW result if the correct flow directions are predefined by the nodalization. The safety relevant, local peak concentrations of hydrogen cannot be resolved by MELCOR. Consequently, the use of the CFD code is indispensable. The correct functioning of the coupling is proven within four steps. At first, the modified MELCOR interface is checked by computing a test case using two

  20. The FLIC conversion codes

    International Nuclear Information System (INIS)

    Basher, J.C.

    1965-05-01

    This report describes the FORTRAN programmes, FLIC 1 and FLIC 2. These programmes convert programmes coded in one dialect of FORTRAN to another dialect of the same language. FLIC 1 is a general pattern recognition and replacement programme whereas FLIC 2 contains extensions directed towards the conversion of FORTRAN II and S2 programmes to EGTRAN 1 - the dialect now in use on the Winfrith KDF9. FII or S2 statements are replaced where possible by their E1 equivalents; other statements which may need changing are flagged. (author)

  1. SPRAY code user's report

    International Nuclear Information System (INIS)

    Shire, P.R.

    1977-03-01

    The SPRAY computer code has been developed to model the effects of postulated sodium spray release from LMFBR piping within containment chambers. The calculation method utilizes gas convection, heat transfer and droplet combustion theory to calculate the pressure and temperature effects within the enclosure. The applicable range is 0-21 mol percent oxygen and .02-.30 inch droplets with or without humidity. Droplet motion and large sodium surface area combine to produce rapid heat release and pressure rise within the enclosed volume

  2. The FLIC conversion codes

    Energy Technology Data Exchange (ETDEWEB)

    Basher, J C [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1965-05-15

    This report describes the FORTRAN programmes, FLIC 1 and FLIC 2. These programmes convert programmes coded in one dialect of FORTRAN to another dialect of the same language. FLIC 1 is a general pattern recognition and replacement programme whereas FLIC 2 contains extensions directed towards the conversion of FORTRAN II and S2 programmes to EGTRAN 1 - the dialect now in use on the Winfrith KDF9. FII or S2 statements are replaced where possible by their E1 equivalents; other statements which may need changing are flagged. (author)

  3. Dynamic Shannon Coding

    OpenAIRE

    Gagie, Travis

    2005-01-01

    We present a new algorithm for dynamic prefix-free coding, based on Shannon coding. We give a simple analysis and prove a better upper bound on the length of the encoding produced than the corresponding bound for dynamic Huffman coding. We show how our algorithm can be modified for efficient length-restricted coding, alphabetic coding and coding with unequal letter costs.

  4. Fundamentals of convolutional coding

    CERN Document Server

    Johannesson, Rolf

    2015-01-01

    Fundamentals of Convolutional Coding, Second Edition, regarded as a bible of convolutional coding brings you a clear and comprehensive discussion of the basic principles of this field * Two new chapters on low-density parity-check (LDPC) convolutional codes and iterative coding * Viterbi, BCJR, BEAST, list, and sequential decoding of convolutional codes * Distance properties of convolutional codes * Includes a downloadable solutions manual

  5. Codes Over Hyperfields

    Directory of Open Access Journals (Sweden)

    Atamewoue Surdive

    2017-12-01

    Full Text Available In this paper, we define linear codes and cyclic codes over a finite Krasner hyperfield and we characterize these codes by their generator matrices and parity check matrices. We also demonstrate that codes over finite Krasner hyperfields are more interesting for code theory than codes over classical finite fields.

  6. Shielding container

    International Nuclear Information System (INIS)

    Darling, K.A.M.

    1981-01-01

    A shielding container incorporates a dense shield, for example of depleted uranium, cast around a tubular member of curvilinear configuration for accommodating a radiation source capsule. A lining for the tubular member, in the form of a close-coiled flexible guide, provides easy replaceability to counter wear while the container is in service. Container life is extended, and maintenance costs are reduced. (author)

  7. German research reactor back-end provisions

    International Nuclear Information System (INIS)

    Koester, Siegfried; Gruber, Gerhard

    2002-01-01

    Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

  8. The Dividend Policy of German Firms

    NARCIS (Netherlands)

    Andres, C.; Betzer, A.; Goergen, M.; Renneboog, L.D.R.

    2008-01-01

    Abstract: German firms pay out a lower proportion of their cash flows, but a higher proportion of their published profits than UK and US firms. We estimate partial adjustment models and report two major findings. First, German firms base their dividend decisions on cash flows rather than published

  9. The Dividend Policy of German Firms

    NARCIS (Netherlands)

    Andres, C.; Betzer, A.; Goergen, M.; Renneboog, L.D.R.

    2008-01-01

    German firms pay out a lower proportion of their cash flows, but a higher proportion of their published profits than UK and US firms. We estimate partial adjustment models and report two major findings. First, German firms base their dividend decisions on cash flows rather than published earnings as

  10. German Schools Abroad: Hotspots of Elite Multilingualism?

    Science.gov (United States)

    Sander, Anne E; Admiraal, Wilfried

    2016-01-01

    While multilingualism itself is a widely analyzed topic, a study about multilingualism at German schools abroad is so far unique. This quantitative study investigates the differences in the size of German expressive and receptive vocabulary between monolingual and multilingual students, aged between 5 and 11 years. A cohort of 65 multilingual…

  11. Massive job cuts threaten East German science

    CERN Multimedia

    Hamer, M

    1990-01-01

    German reunification could result in thousands of scientists losing their jobs. At the end of this year the East German state budget for science will run out. Scientists in the East are keen to find Western support to protect their research (1 page).

  12. When do German Firms Change their Dividends?

    NARCIS (Netherlands)

    Correia Da Silva, L.; Goergen, M.; Renneboog, L.D.R.

    2002-01-01

    Anecdotal evidence suggests that the dividend policy of German firms is more flexible than the one of their Anglo-American counterparts.This paper analyses the decision to change the dividend for a panel of 221 German firms from 1984 to 1994.The choice of the period of study is motivated by the fact

  13. Lexical Reading in Dysfluent Readers of German

    Science.gov (United States)

    Gangl, Melanie; Moll, Kristina; Jones, Manon W.; Banfi, Chiara; Schulte-Körne, Gerd; Landerl, Karin

    2018-01-01

    Dyslexia in consistent orthographies like German is characterized by dysfluent reading, which is often assumed to result from failure to build up an orthographic lexicon and overreliance on decoding. However, earlier evidence indicates effects of lexical processing at least in some German dyslexic readers. We investigated variations in reading…

  14. Facebook Used in a German Film Project

    Science.gov (United States)

    Leier, Vera

    2011-01-01

    Looking for a way to make German language study more relevant and to step out of the conventional classroom setting, I introduced Facebook (FB) as a learning platform to my intermediate German students at the University of Canterbury, New Zealand. The students took part in a film competition. A FB group was created and the films were uploaded. The…

  15. DIMA – Annotation guidelines for German intonation

    DEFF Research Database (Denmark)

    Kügler, Frank; Smolibocki, Bernadett; Arnold, Denis

    2015-01-01

    This paper presents newly developed guidelines for prosodic annotation of German as a consensus system agreed upon by German intonologists. The DIMA system is rooted in the framework of autosegmental-metrical phonology. One important goal of the consensus is to make exchanging data between groups...

  16. Teaching German Culture: An Alternative Approach.

    Science.gov (United States)

    Ray, Maruta L.

    1985-01-01

    Describes a college course on German culture in which the criterion for the inclusion of any topic in the syllabus is its mention--preferably recurrent--in the German press. Additional emphasis is placed upon the historical background of the current events. Classes are a combination of films, lectures, discussions, and student reports. (SED)

  17. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  18. [Hand surgery in the German DRG System 2007].

    Science.gov (United States)

    Franz, D; Windolf, J; Kaufmann, M; Siebert, C H; Roeder, N

    2007-05-01

    Hand surgery often needs only a short length of stay in hospital. Patients' comorbidity is low. Many hand surgery procedures do not need inpatient structures. Up until 2006 special procedures of hand surgery could not be coded. The DRG structure did not separate very complex and less complex operations. Specialized hospitals needed a proper case allocation of their patients within the G-DRG system. The DRG structure concerning hand surgery increased in version 2007 of the G-DRG system. The main parameter of DRG splitting is the complexity of the operation. Furthermore additional criteria such as more than one significant OR procedure, the patients' age, or special diagnoses influence case allocation. A special OPS code for complex cases treated with hand surgery was implemented. The changes in the DRG structure and the implementation of the new OPS code for complex cases establish a strong basis for the identification of different patient costs. Different case allocation leads to different economic impacts on departments of hand surgery. Whether the new OPS code becomes a DRG splitting parameter has to be calculated by the German DRG Institute for further DRG versions.

  19. German versus Nordic Board Models

    DEFF Research Database (Denmark)

    Ringe, Georg

    2016-01-01

    Board structure is an important component of the individual governance of firms, and the appropriateness of the various models is one of the most debated issues in corporate governance today. A comparison of the Nordic and German approaches to the structure of corporate boards reveals stark...... conceptual differences, as emphasized by the 2014 Lekvall Report on the Nordic Corporate Governance Model. This article provides a conceptual comparison between the two approaches to board structure and confirms the fundamental divergence between both models. However, relying on a number of recent legal...... changes and developments in business practice, the article argues that board practices in the two systems effectively blur the structural distinction, and that board organization is converging in practice. It thereby contributes to the broader debates on functionality and comparative corporate law...

  20. German standard problem No. 2

    International Nuclear Information System (INIS)

    Burkhardt, R.

    1980-02-01

    The German Standard Problem Nr. 2 (primary circuits) is meant to check whether the presently available computing programs dealing with ECCS problems are suitable to reflect with sufficient accuracy reload and flooding processes. Changing from conventional calculation methods to the ''best-estimate'' method requires for possibility of exact comparison, as is the case here because of experimental results from the primary circuit test plant. The test plant of KWU Erlangen with primary circuit modeups on a 1:134 scale with exact level indications allows comparative testing where emergency cooling water is loaded into the system filled with saturated steam over cold lanes, or rather over the annulus modeup. The report on hand goes into detail about calculations, anticipated results and their comparison to experimental results. (orig./RW) [de

  1. German General Staff Officer Education and Current Challenges

    National Research Council Canada - National Science Library

    Groeters, Thomas

    2006-01-01

    "German General Staff Officer Education and Current Challenges" examines the institutional education of German General Staff Officers, as experienced by the author, and offers a "Conceptual Competency...

  2. [Orthopedic and trauma surgery in the German DRG system. Recent developments].

    Science.gov (United States)

    Franz, D; Schemmann, F; Selter, D D; Wirtz, D C; Roeder, N; Siebert, H; Mahlke, L

    2012-07-01

    Orthopedics and trauma surgery are subject to continuous medical advancement. The correct and performance-based case allocation by German diagnosis-related groups (G-DRG) is a major challenge. This article analyzes and assesses current developments in orthopedics and trauma surgery in the areas of coding of diagnoses and medical procedures and the development of the 2012 G-DRG system. The relevant diagnoses, medical procedures and G-DRGs in the versions 2011 and 2012 were analyzed based on the publications of the German DRG Institute (InEK) and the German Institute of Medical Documentation and Information (DIMDI). Changes were made for the International Classification of Diseases (ICD) coding of complex cases with medical complications, the procedure coding for spinal surgery and for hand and foot surgery. The G-DRG structures were modified for endoprosthetic surgery on ankle, shoulder and elbow joints. The definition of modular structured endoprostheses was clarified. The G-DRG system for orthopedic and trauma surgery appears to be largely consolidated. The current phase of the evolution of the G-DRG system is primarily aimed at developing most exact descriptions and definitions of the content and mutual delimitation of operation and procedures coding (OPS). This is an essential prerequisite for a correct and performance-based case allocation in the G-DRG system.

  3. Urban Green Infrastructure: German Experience

    Directory of Open Access Journals (Sweden)

    Diana Olegovna Dushkova

    2016-06-01

    Full Text Available The paper presents a concept of urban green infrastructure and analyzes the features of its implementation in the urban development programmes of German cities. We analyzed the most shared articles devoted to the urban green infrastructure to see different approaches to definition of this term. It is based on materials of field research in the cities of Berlin and Leipzig in 2014-2015, international and national scientific publications. During the process of preparing the paper, consultations have been held with experts from scientific institutions and Administrations of Berlin and Leipzig as well as local experts from environmental organizations of both cities. Using the German cities of Berlin and Leipzig as examples, this paper identifies how the concept can be implemented in the program of urban development. It presents the main elements of green city model, which include mitigation of negative anthropogenic impact on the environment under the framework of urban sustainable development. Essential part of it is a complex ecological policy as a major necessary tool for the implementation of the green urban infrastructure concept. This ecological policy should embody not only some ecological measurements, but also a greening of all urban infrastructure elements as well as implementation of sustainable living with a greater awareness of the resources, which are used in everyday life, and development of environmental thinking among urban citizens. Urban green infrastructure is a unity of four main components: green building, green transportation, eco-friendly waste management, green transport routes and ecological corridors. Experience in the development of urban green infrastructure in Germany can be useful to improve the environmental situation in Russian cities.

  4. Vector Network Coding Algorithms

    OpenAIRE

    Ebrahimi, Javad; Fragouli, Christina

    2010-01-01

    We develop new algebraic algorithms for scalar and vector network coding. In vector network coding, the source multicasts information by transmitting vectors of length L, while intermediate nodes process and combine their incoming packets by multiplying them with L x L coding matrices that play a similar role as coding c in scalar coding. Our algorithms for scalar network jointly optimize the employed field size while selecting the coding coefficients. Similarly, for vector coding, our algori...

  5. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  6. A container

    DEFF Research Database (Denmark)

    2012-01-01

    A container assembly for the containment of fluids or solids under a pressure different from the ambient pressure comprising a container (2) comprising an opening and an annular sealing, a lid (3) comprising a central portion (5) and engagement means (7) for engaging the annular flange, and sealing...... means (10) wherein the engagement means (7) is adapted, via the sealing means, to seal the opening when the pressure of the container assembly differs from the ambient pressure in such a way that the central portion (5) flexes in the axial direction which leads to a radial tightening of the engagement...... means (7) to the container, wherein the container further comprises locking means (12) that can be positioned so that the central portion is hindered from flexing in at least one direction....

  7. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

    International Nuclear Information System (INIS)

    Huang, Xi

    2016-07-01

    In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

  8. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xi

    2016-07-15

    In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

  9. Shielded container

    International Nuclear Information System (INIS)

    Fries, B.A.

    1978-01-01

    A shielded container for transportation of radioactive materials is disclosed in which leakage from the container is minimized due to constructional features including, inter alia, forming the container of a series of telescoping members having sliding fits between adjacent side walls and having at least two of the members including machine sealed lids and at least two of the elements including hand-tightenable caps

  10. Multimedia signal coding and transmission

    CERN Document Server

    Ohm, Jens-Rainer

    2015-01-01

    This textbook covers the theoretical background of one- and multidimensional signal processing, statistical analysis and modelling, coding and information theory with regard to the principles and design of image, video and audio compression systems. The theoretical concepts are augmented by practical examples of algorithms for multimedia signal coding technology, and related transmission aspects. On this basis, principles behind multimedia coding standards, including most recent developments like High Efficiency Video Coding, can be well understood. Furthermore, potential advances in future development are pointed out. Numerous figures and examples help to illustrate the concepts covered. The book was developed on the basis of a graduate-level university course, and most chapters are supplemented by exercises. The book is also a self-contained introduction both for researchers and developers of multimedia compression systems in industry.

  11. Homological stabilizer codes

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Jonas T., E-mail: jonastyleranderson@gmail.com

    2013-03-15

    In this paper we define homological stabilizer codes on qubits which encompass codes such as Kitaev's toric code and the topological color codes. These codes are defined solely by the graphs they reside on. This feature allows us to use properties of topological graph theory to determine the graphs which are suitable as homological stabilizer codes. We then show that all toric codes are equivalent to homological stabilizer codes on 4-valent graphs. We show that the topological color codes and toric codes correspond to two distinct classes of graphs. We define the notion of label set equivalencies and show that under a small set of constraints the only homological stabilizer codes without local logical operators are equivalent to Kitaev's toric code or to the topological color codes. - Highlights: Black-Right-Pointing-Pointer We show that Kitaev's toric codes are equivalent to homological stabilizer codes on 4-valent graphs. Black-Right-Pointing-Pointer We show that toric codes and color codes correspond to homological stabilizer codes on distinct graphs. Black-Right-Pointing-Pointer We find and classify all 2D homological stabilizer codes. Black-Right-Pointing-Pointer We find optimal codes among the homological stabilizer codes.

  12. The ZPIC educational code suite

    Science.gov (United States)

    Calado, R.; Pardal, M.; Ninhos, P.; Helm, A.; Mori, W. B.; Decyk, V. K.; Vieira, J.; Silva, L. O.; Fonseca, R. A.

    2017-10-01

    Particle-in-Cell (PIC) codes are used in almost all areas of plasma physics, such as fusion energy research, plasma accelerators, space physics, ion propulsion, and plasma processing, and many other areas. In this work, we present the ZPIC educational code suite, a new initiative to foster training in plasma physics using computer simulations. Leveraging on our expertise and experience from the development and use of the OSIRIS PIC code, we have developed a suite of 1D/2D fully relativistic electromagnetic PIC codes, as well as 1D electrostatic. These codes are self-contained and require only a standard laptop/desktop computer with a C compiler to be run. The output files are written in a new file format called ZDF that can be easily read using the supplied routines in a number of languages, such as Python, and IDL. The code suite also includes a number of example problems that can be used to illustrate several textbook and advanced plasma mechanisms, including instructions for parameter space exploration. We also invite contributions to this repository of test problems that will be made freely available to the community provided the input files comply with the format defined by the ZPIC team. The code suite is freely available and hosted on GitHub at https://github.com/zambzamb/zpic. Work partially supported by PICKSC.

  13. LFSC - Linac Feedback Simulation Code

    International Nuclear Information System (INIS)

    Ivanov, Valentin; Fermilab

    2008-01-01

    The computer program LFSC ( ) is a numerical tool for simulation beam based feedback in high performance linacs. The code LFSC is based on the earlier version developed by a collective of authors at SLAC (L.Hendrickson, R. McEwen, T. Himel, H. Shoaee, S. Shah, P. Emma, P. Schultz) during 1990-2005. That code was successively used in simulation of SLC, TESLA, CLIC and NLC projects. It can simulate as pulse-to-pulse feedback on timescale corresponding to 5-100 Hz, as slower feedbacks, operating in the 0.1-1 Hz range in the Main Linac and Beam Delivery System. The code LFSC is running under Matlab for MS Windows operating system. It contains about 30,000 lines of source code in more than 260 subroutines. The code uses the LIAR ('Linear Accelerator Research code') for particle tracking under ground motion and technical noise perturbations. It uses the Guinea Pig code to simulate the luminosity performance. A set of input files includes the lattice description (XSIF format), and plane text files with numerical parameters, wake fields, ground motion data etc. The Matlab environment provides a flexible system for graphical output

  14. Basic principles and results of the German risk study

    International Nuclear Information System (INIS)

    Heuser, F.W.; Bayer, A.

    1980-01-01

    In June 1976 the Federal Ministry for Research and Technology had commissioned the Gesellschaft fuer Reaktorsicherheit to write the German Risk Study, the first part of which has now been completed after three years of work and has been publicized recently. The German Risk Study is an attempt to define the societal risk posed by accidents in nuclear power plants under conditions in Germany. For this purpose, the accident rates and the resultant health hazards were determined. By adopting most of the basic premises and methods of the American Rasmussen Study, the German study is to allow a comparison to be made with the results of that study. The calculations were based on 19 sites with a total of 25 nuclear generating units presently in operation, under construction or in the licensing procedure in the Federal Republic of Germany. The technical studies were conducted on a 1300 MW PWR as the representative example. The results show that the decisive contributions are made by uncontrolled minor loss-of-coolant accidents and by failures of power supply (emergency power case). Large loss-of-coolant accidents do not play a role. The study also shows the decisive safety function of the containment. (orig.) [de

  15. The German energetic future, comparison with the France Negatep scenario

    International Nuclear Information System (INIS)

    Acket, C.; Bacher, P.

    2011-01-01

    As the Germans have decided to abandon nuclear energy, which today provides 23 % of their electricity, while fossil fuels provide 58%, the authors aim at answering two important questions. The first one is whether it is possible to cope without the non carbon nuclear energy while simultaneously reducing the CO 2 emissions. Considering the current level of German CO 2 emissions (over 9 tonnes per year per person), while the objective is to reach less than 2 tonnes per year by 2050, the second question is whether energy efficiency and renewable energies can be the solution. The authors present several scenarios meeting the overall emission objectives (a scenario dividing by two CO 2 emissions between 2008 and 2050, and eight scenarios aiming at five times less emissions in 2050 than in 2008), with different transition periods for nuclear energy. Since in all the scenarios, there is no nuclear left in 2050, they examine the energy balance in 2050 and point out the main characteristics of the German energy mix at that time. Almost identical with another document with the same title, this version contains figures which are not present in the other one

  16. [ENT and head and neck surgery in the German DRG system 2007].

    Science.gov (United States)

    Franz, D; Roeder, N; Hörmann, K; Alberty, J

    2007-07-01

    The German DRG system has been further developed into version 2007. For ENT and head and neck surgery, significant changes in the coding of diagnoses and medical operations as well as in the the DRG structure have been made. New ICD codes for sleep apnoea and acquired tracheal stenosis have been implemented. Surgery on the acoustic meatus, removal of auricle hyaline cartilage for transplantation (e. g. rhinosurgery) and tonsillotomy have been coded in the 2007 version. In addition, the DRG structure has been improved. Case allocation of more than one significant operation has been established. The G-DRG system has gained in complexity. High demands are made on the coding of complex cases, whereas standard cases require mostly only one specific diagnosis and one specific OPS code. The quality of case allocation for ENT patients within the G-DRG system has been improved. Nevertheless, further adjustments of the G-DRG system are necessary.

  17. MAD parsing and conversion code

    International Nuclear Information System (INIS)

    Mokhov, Dmitri N.

    2000-01-01

    The authors describe design and implementation issues while developing an embeddable MAD language parser. Two working applications of the parser are also described, namely, MAD-> C++ converter and C++ factory. The report contains some relevant details about the parser and examples of converted code. It also describes some of the problems that were encountered and the solutions found for them

  18. [The boycott against German scientists and the German language after World War I].

    Science.gov (United States)

    Reinbothe, R

    2013-12-01

    After the First World War, the Allied academies of sciences staged a boycott against German scientists and the German language. The objective of the boycott was to prevent the re-establishment of the prewar dominance of German scientists, the German language and German publications in the area of international scientific cooperation. Therefore the Allies excluded German scientists and the German language from international associations, congresses and publications, while they created new international scientific organizations under their leadership. Medical associations and congresses were also affected, e. g. congresses on surgery, ophthalmology and tuberculosis. Allied physicians replaced the "International Anti-Tuberculosis Association" founded in Berlin in 1902 with the "Union Internationale contre la Tuberculose"/"International Union against Tuberculosis", founded in Paris in 1920. Only French and English were used as the official languages of the new scientific organizations, just as in the League of Nations. The boycott was based on the fact that the German scientists had denied German war guilt and war crimes and glorified German militarism in a manifesto "To The Civilized World!" in 1914. The boycott first started in 1919 and had to be abolished in 1926, when Germany became a member of the League of Nations. Many German and foreign physicians as well as other scientists protested against the boycott. Some German scientists and institutions even staged a counter-boycott impeding the resumption of international collaboration. The boycott entailed an enduring decline of German as an international scientific language. After the Second World War scientists of the victorious Western Powers implemented a complete reorganization of the international scientific arena, based on the same organizational structures and language restrictions they had built up in 1919/1920. At the same time scientists from the U.S.A. staged an active language and publication policy, in

  19. Diagnostic Coding for Epilepsy.

    Science.gov (United States)

    Williams, Korwyn; Nuwer, Marc R; Buchhalter, Jeffrey R

    2016-02-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  20. Coding of Neuroinfectious Diseases.

    Science.gov (United States)

    Barkley, Gregory L

    2015-12-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  1. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  2. German energy market 2014; Deutscher Energiemarkt 2014

    Energy Technology Data Exchange (ETDEWEB)

    Schiffer, Hans-Wilhelm [World Energy Council, London (United Kingdom). World Energy Resources; Weltenergierat - Deutschland, Berlin (Germany). Arbeitsgruppe ' Energie fuer Deutschland'

    2015-03-15

    In 2014 the German government's primary goal of engaging German power suppliers to step up their production of renewable energy while speeding up energy efficiency improvement measures continued to dominate the debate. The present article provides an updated overview of the German energy market. Following on from last year's edition it gives a condensed synopsis of key indicators of the energy economy. Besides summarising general facts about the energy mix it goes into detail about the following individual energy resources: crude oil, natural gas, brown coal, hard coal, nuclear energy and renewable energies. It also explains current price trends in both the international and domestic markets.

  3. Comparison between Dutch and German buildings

    Energy Technology Data Exchange (ETDEWEB)

    Lony, R.J.M.; Molenaar, D.J.; Rietkerk, J.; Schuiling, D.J.B.W.; Zeiler, W. [TU/e, Univ. of Technology Eindhoven (Netherlands); Brunk, M. [RWTH Aachen (Germany)

    2006-07-01

    German buildings are often seen as an example to Dutch architects and Dutch building services consultants. Goal of this article is to examine and to understand differences between the Dutch and German top office buildings. Objective is to examine to which extent these buildings were designed intelligently. An Intelligent Building is one that provides a productive cost effective environment through the optimisation of six basic elements; site, skin, systems, structures, services, space plan and staff and the interrelationship between them. Based on these six aspects the comparison is made between Dutch and German buildings. (orig.)

  4. German causative events with placement verbs

    Directory of Open Access Journals (Sweden)

    De Knop Sabine

    2016-06-01

    Full Text Available Several studies have described the semantic uses of German posture verbs, but only few have dealt with German placement verbs. The present study wants to make up for this gap. Starting from a collection of examples from the core corpora of the Digitales Wörterbuch der Deutschen Sprache (DWDS and some former studies on posture verbs, it first describes the variety of the most common German placement verbs stellen (‘to put upright’, legen (‘to lay down’, setzen (‘to set’ and stecken (‘to stick’.

  5. Cost reduction potentials in the German market for balancing power

    International Nuclear Information System (INIS)

    Flinkerbusch, Kai; Heuterkes, Michael

    2010-01-01

    This article examines potential cost reductions in the market for balancing power by pooling all four German control areas. In a united control area both the procurement and the production of balancing power may be more efficient than in four separated control areas. Our data contain bids on energy procurement as well as balancing power flows in the period from December 2007 to November 2008. A reference scenario simulates the market results for secondary and tertiary balancing power. Subsequently, we simulate a united control area. We show that in the period under review the total costs of balancing power are reduced by 17%. (author)

  6. The German scientific balloon and sounding rocket projects

    International Nuclear Information System (INIS)

    Dalh, A.F.

    1978-01-01

    This report contains information on the sounding rocket projects: experiment preparation for spacelab (astronomy), aeronomy, magnetosphere, and material science. Except for material science the scientific balloon projects are performed in the some scientific fields, but with a strong emphasis on astronomical research. It is tried to provide by means of tables a survey as complete as possible of the projects for the time since the last symposium in Elmau and of the plans for the future until 1981. The scientific balloon and sounding rocket projects form a small succesful part of the German space research programme. (author)

  7. The German scientific balloon and sounding rocket programme

    International Nuclear Information System (INIS)

    Dahl, A.F.

    1980-01-01

    This report contains information on sounding rocket projects in the scientific field of astronomy, aeronomy, magnetosphere, and material science under microgravity. The scientific balloon projects are performed with emphasis on astronomical research. By means of tables it is attempted to give a survey, as complete as possible, of the projects the time since the last symposium in Ajaccio, Corsica, and of preparations and plans for the future until 1983. The scientific balloon and sounding rocket projects form a small successful part of the German space research programme. (Auth.)

  8. Main results of the German risk study - phase B

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1987-01-01

    To start the author introduces briefly some comments on main tasks and objectives of risk analysises which at least after the Chernobyl accident should be emphasized more explicitly. Following on some main results of the system - and accident event tree analysis of the German Risk Study, Phase B, are summarized. The second part of this paper deals with the analysis of core melt accidents performed in context of the study. Hetero investigations on the hydrogen problem and investigations on containment venting after a core melt accident will be discussed in more detail

  9. Objectives and present status of the German risk evaluation study

    International Nuclear Information System (INIS)

    Birkhofer, A.; Koeberlein, K.; Heuser, F.W.

    1977-01-01

    For the German risk evaluation study, analogous to the Rasmussen report (WASH--1400), embarked upon in June 1976, the Kernkraftwerk Biblis B serves as the plant of reference. The first interim results are available for various sub-headings of the study. The main finding seems to be the decisive importance of the containment in limiting the accident consequences even in those cases where, on account of postulated failure of safety systems, the melt down of the reactor core is to be expected. (orig./HP) [de

  10. Illicit operation of industrial plant subject to licensing, or of other installations within the purview of the German Federal Emission Control Act (BImSchG), which have been shut down for protection against hazards (section 327, subsection 2, No. 1 Penal Code (StGB)); Das unerlaubte Betreiben von genehmigungsbeduerftigen Anlagen oder sonstigen Anlagen im Sinne des Bundes-Immissionsschutzgesetzes, deren Betrieb zum Schutz vor Gefahren untersagt worden ist (Paragraph 327 Abs.2 Nr.1 StGB)

    Energy Technology Data Exchange (ETDEWEB)

    Ocker, A.

    1995-12-31

    The 18th act of 28 March 1980 for amendment of the German Criminal Code (StGB) incorporated the provisions governing the criminal offence of illicit operation of installations subject to licensing into the StGB. These provisions have until then been forming part of the BImSchG (Federal Act on Emission Control). The study in hand presents a discussion of section 327, subsection 2, No. 1 StGB, because this provision represents a fundamental type of an administration accessory criminal offence and thus is suitable to be taken as a basis for an analysis of the scope of problems covered by the StGB, but having an effect on and being interlaced with offences governed by other acts and legal provisions. The study addresses inter alia items such as the object of legal protection defined by this section of the StGB, the provisions defining the licensability of a non-licensed installation in operation, and the consequences of defective decisions under administrative law on the applicability of criminal law provisions. The specific aspects of section 327 StGB, which are of a dominantly administrative nature, are discussed, in particular those referring to the definition of the term ``industrial installation`` as defined by the BImSchG. [Deutsch] Durch das 18. Strafrechtsaenderungsgesetz vom 28.3.1980 wurde der Straftatbestand des unerlaubten Betreibens von Anlagen, die einer Genehmigung nach dem Bundesimmissionsschutzgesetz beduerfen, aus dem BImSchG in das Kernstrafrecht ueberfuehrt. Die vorliegende Untersuchung versucht eine eingehendere Auseinandersetzung mit Para. 327 Abs. 2 Nr. 1 StGB, weil diese Vorschrift als ein Grundtypus der verwaltungsakzessorischen Straftatbestaende die Gelegenheiit zur vertieften Diskussion von umfassenderen strafrechtlichen Problemkreisen gibt. Angesprochen sind insoweit vor allem die Frage des geschuetzten Rechtsgutes der Norm, der Behandlung der materiellen Genehmigungsfaehigkeit eines ungenehmigten Anlagenbetriebs sowie der Auswirkungen von

  11. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  12. Sharps container

    Science.gov (United States)

    Lee, Angelene M. (Inventor)

    1992-01-01

    This invention relates to a system for use in disposing of potentially hazardous items and more particularly a Sharps receptacle for used hypodermic needles and the like. A Sharps container is constructed from lightweight alodined nonmagnetic metal material with a cup member having an elongated tapered shape and length greater than its transverse dimensions. A magnet in the cup member provides for metal retention in the container. A nonmagnetic lid member has an opening and spring biased closure flap member. The flap member is constructed from stainless steel. A Velcro patch on the container permits selective attachment at desired locations.

  13. Loads on EPR containment after RPV failure at high pressure

    International Nuclear Information System (INIS)

    Jacobs, G.

    1995-01-01

    As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

  14. Vector Network Coding

    OpenAIRE

    Ebrahimi, Javad; Fragouli, Christina

    2010-01-01

    We develop new algebraic algorithms for scalar and vector network coding. In vector network coding, the source multicasts information by transmitting vectors of length L, while intermediate nodes process and combine their incoming packets by multiplying them with L X L coding matrices that play a similar role as coding coefficients in scalar coding. Our algorithms for scalar network jointly optimize the employed field size while selecting the coding coefficients. Similarly, for vector co...

  15. Entropy Coding in HEVC

    OpenAIRE

    Sze, Vivienne; Marpe, Detlev

    2014-01-01

    Context-Based Adaptive Binary Arithmetic Coding (CABAC) is a method of entropy coding first introduced in H.264/AVC and now used in the latest High Efficiency Video Coding (HEVC) standard. While it provides high coding efficiency, the data dependencies in H.264/AVC CABAC make it challenging to parallelize and thus limit its throughput. Accordingly, during the standardization of entropy coding for HEVC, both aspects of coding efficiency and throughput were considered. This chapter describes th...

  16. Generalized concatenated quantum codes

    International Nuclear Information System (INIS)

    Grassl, Markus; Shor, Peter; Smith, Graeme; Smolin, John; Zeng Bei

    2009-01-01

    We discuss the concept of generalized concatenated quantum codes. This generalized concatenation method provides a systematical way for constructing good quantum codes, both stabilizer codes and nonadditive codes. Using this method, we construct families of single-error-correcting nonadditive quantum codes, in both binary and nonbinary cases, which not only outperform any stabilizer codes for finite block length but also asymptotically meet the quantum Hamming bound for large block length.

  17. Rateless feedback codes

    DEFF Research Database (Denmark)

    Sørensen, Jesper Hemming; Koike-Akino, Toshiaki; Orlik, Philip

    2012-01-01

    This paper proposes a concept called rateless feedback coding. We redesign the existing LT and Raptor codes, by introducing new degree distributions for the case when a few feedback opportunities are available. We show that incorporating feedback to LT codes can significantly decrease both...... the coding overhead and the encoding/decoding complexity. Moreover, we show that, at the price of a slight increase in the coding overhead, linear complexity is achieved with Raptor feedback coding....

  18. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  19. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

    This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV

  20. Coding for dummies

    CERN Document Server

    Abraham, Nikhil

    2015-01-01

    Hands-on exercises help you learn to code like a pro No coding experience is required for Coding For Dummies,your one-stop guide to building a foundation of knowledge inwriting computer code for web, application, and softwaredevelopment. It doesn't matter if you've dabbled in coding or neverwritten a line of code, this book guides you through the basics.Using foundational web development languages like HTML, CSS, andJavaScript, it explains in plain English how coding works and whyit's needed. Online exercises developed by Codecademy, a leading online codetraining site, help hone coding skill

  1. [Coding in general practice-Will the ICD-11 be a step forward?

    Science.gov (United States)

    Kühlein, Thomas; Virtanen, Martti; Claus, Christoph; Popert, Uwe; van Boven, Kees

    2018-07-01

    Primary care physicians in Germany don't benefit from coding diagnoses-they are coding for the needs of others. For coding, they mostly are using either the thesaurus of the German Institute of Medical Documentation and Information (DIMDI) or self-made cheat-sheets. Coding quality is low but seems to be sufficient for the main use case of the resulting data, which is the morbidity adjusted risk compensation scheme that distributes financial resources between the many German health insurance companies.Neither the International Classification of Diseases and Health Related Problems (ICD-10) nor the German thesaurus as an interface terminology are adequate for coding in primary care. The ICD-11 itself will not recognizably be a step forward from the perspective of primary care. At least the browser database format will be advantageous. An implementation into the 182 different electronic health records (EHR) on the German market would probably standardize the coding process and make code finding easier. This method of coding would still be more cumbersome than the current coding with self-made cheat-sheets.The first steps towards a useful official cheat-sheet for primary care have been taken, awaiting implementation and evaluation. The International Classification of Primary Care (ICPC-2) already provides an adequate classification standard for primary care that can also be used in combination with ICD-10. A new version of ICPC (ICPC-3) is under development. As the ICPC-2 has already been integrated into the foundation layer of ICD-11 it might easily become the future standard for coding in primary care. Improving communication between the different EHR would make taking over codes from other healthcare providers possible. Another opportunity to improve the coding quality might be creating use cases for the resulting data for the primary care physicians themselves.

  2. African Americans Who Teach German Language and Culture.

    Science.gov (United States)

    Fikes, Robert Jr.

    2001-01-01

    A large number of black scholars have pursued advanced degrees in the German language, history, and culture. Describes the history of African American interest in the German language and culture, highlighting various black scholars who have studied German over the years. Presents data on African Americans in German graduate programs and examines…

  3. Crossing the Lexicon: Anglicisms in the German Hip Hop Community

    Science.gov (United States)

    Garley, Matthew E.

    2012-01-01

    The influence of English on German has been an ongoing subject of intense popular and academic interest in the German sphere. In order to better understand this language contact situation, this research project investigates anglicisms--instances of English language material in a German language context--in the German hip hop community, where the…

  4. USA: German in the Changing Landscape of Postsecondary Education

    Science.gov (United States)

    Tatlock, Lynne

    2010-01-01

    This article identifies recent indicators of the state of German Studies in the United States with special attention to postsecondary enrollments in German. It additionally reviews challenges to the postsecondary teaching of German as they manifest themselves both locally and nationally, including the positioning of German Studies in the life of…

  5. A German format for pupils’ training

    CERN Multimedia

    Antonella Del Rosso

    2012-01-01

    Every year CERN welcomes thousands of pupils from schools worldwide for a half-day visit to the Laboratory. However, since 2011 about ten selected students from Germany have been given the opportunity to experience CERN in much greater depth. They are fully sponsored by the German Ministry of Education and supported by an organising structure at TU Dresden - the Dresden University of Technology - led by Michael Kobel. It’s an investment that's paying off in Germany.   The German teachers who participated in the “Netzwerk Teilchenwelt” project, at CERN last week. “Netzwerk Teilchenwelt” is a project that involves 23 German universities, the DESY Laboratory, several schools and, of course, CERN. Launched in 2010 with a contribution from the German Ministry for Science and Research of about 1 million euros over three years, the project has so far involved over 4,000 students and 500 teachers. “Thanks to this project, both pupils...

  6. NPPCI - topics in the German Democratic Republic

    International Nuclear Information System (INIS)

    Ziegenbein, D.

    1986-01-01

    This paper summarizes research and development activities in the field of computerized operator support systems, self-powered detectors, boiling diagnostic and loose part detection systems in the German Democratic Republic

  7. The German Physical Society Under National Socialism

    Science.gov (United States)

    Hoffmann, Dieter; Walker, Mark

    2004-12-01

    The history of the German Physical Society from 1933 to 1945 is not the same as a comprehensive history of physics under Adolf Hitler, but it does reflect important aspects of physicists' work and life during the Third Reich.

  8. Approaching German Culture: A Tentative Analysis

    Science.gov (United States)

    Tinsley, Royal; Woloshin, David

    1974-01-01

    A comparative analysis of the five universal problems of cultural orientation: 1) human nature, 2) social relations, 3) man and nature, 4) time, 5) space, as they are reflected in German and American culture. (PP)

  9. German approach to VLLW management

    International Nuclear Information System (INIS)

    Broecking, D.

    1997-01-01

    Waste generated in German nuclear facilities is exhaustively and selectively manage through a system adapted to the waste's characteristics. The management of the waste ensures that the impact on the workers, population and the environment is acceptable. This is done through a detailed documentation and quality assurance program which applies not only to radioactive waste but also to cleanable material used within a licensed practice. In Germany the producer is responsible for the waste generated in a licensed facility and is therefore responsible for the correct disposal, which depends on the waste's characteristics. Since the waste producer requires a license for all activities involving radioactive substances, the atomic authority is continuously informed and can therefore monitor the producer for compliance with the regulations. The use and disposal of all material in a licensed practice is documented and can therefore be traced by the authorities. Clearance is seen in Germany as the best way of managing non-radioactive material in a licensed practice. Germany has developed clearance procedures which guarantee that after clearance the radiological impact is negligible. (author)

  10. German innovation initiative for nanotechnology

    Science.gov (United States)

    Rieke, Volker; Bachmann, Gerd

    2004-10-01

    In many areas of nanotechnology, Germany can count on a good knowledge basis due to its diverse activities in nanosciences. This knowledge basis, when paired with the production and sales structures needed for implementation and the internationally renowned German talent for system integration, should consequently lead to success in the marketplace. And this is exactly the field of application for the innovation initiative "Nanotechnologie erobert Märkte" (nanotechnology conquers markets) and for the new BMBF strategy in support of nanotechnology. Until now, aspects of nanotechnology have been advanced within the confines of their respective technical subject areas. However, the primary aim of incorporating them into an overall national strategy is to build on Germany's well-developed and internationally competitive research in science and technology to tap the potential of Germany's important industrial sectors for the application of nanotechnology through joint research projects (leading-edge innovations) that strategically target the value-added chain. This development is to be supported by government education policy to remedy a threatening shortage of skilled professionals. To realize that goal, forward-looking political policymaking must become oriented to a uniform concept of innovation, one that takes into consideration all facets of new technological advances that can contribute to a new culture of innovation in Germany. And that includes education and research policy as well as a climate that encourages and supports innovation in science, business and society.

  11. German innovation initiative for nanotechnology

    International Nuclear Information System (INIS)

    Rieke, Volker; Bachmann, Gerd

    2004-01-01

    In many areas of nanotechnology, Germany can count on a good knowledge basis due to its diverse activities in nanosciences. This knowledge basis, when paired with the production and sales structures needed for implementation and the internationally renowned German talent for system integration, should consequently lead to success in the marketplace. And this is exactly the field of application for the innovation initiative 'Nanotechnologie erobert Maerkte' (nanotechnology conquers markets) and for the new BMBF strategy in support of nanotechnology. Until now, aspects of nanotechnology have been advanced within the confines of their respective technical subject areas. However, the primary aim of incorporating them into an overall national strategy is to build on Germany's well-developed and internationally competitive research in science and technology to tap the potential of Germany's important industrial sectors for the application of nanotechnology through joint research projects (leading-edge innovations) that strategically target the value-added chain. This development is to be supported by government education policy to remedy a threatening shortage of skilled professionals. To realize that goal, forward-looking political policymaking must become oriented to a uniform concept of innovation, one that takes into consideration all facets of new technological advances that can contribute to a new culture of innovation in Germany. And that includes education and research policy as well as a climate that encourages and supports innovation in science, business and society

  12. German-Brazilian nuclear deal

    International Nuclear Information System (INIS)

    Krugmann, H.

    1981-01-01

    Examination of the arguments in favor of the nuclear deal with West Germany and the resulting program suggests that revisions of both are in order to make them more compatible with Brazil's national interests. The deficiencies of current policy appear to be too weighty and numerous to be ignored. Sooner or later the government will have to move toward adjusting its nuclear agreement with West Germany, if not for the reasons discussed here then for lack of capital. Current estimates of the nuclear package lie in the range of $25 to $30 billion, compared to an initial projection of about $5 billion. The deal has become so expensive that it would draw capital from the hydropower and alcohol programs essential for the short and medium-term energy needs of the country. Mr. Krugman feels the Brazilian government should hold off on further nuclear contracts. And it should thoroughly reassess what Brazil's nuclear energy and technology requirements are and how to meet them. There are indications that the reassessment process is already underway. As long as the German nuclear industry depends on the sale of technology to Brazil, the Brazilian government will have considerable bargaining power to enforce further changes in the deal. If this power is used wisely, the result could be cooperation between the two countries toward nuclear options that are consistent with Brazil's energy and development needs

  13. Franco-German nuclear cooperation

    International Nuclear Information System (INIS)

    Leny, J.C.; Huettl, A.

    1996-01-01

    Nuclear energy is the number one power source in the European Union. However, the first generation units would be replaced from the year 2010 onwards. In this prospect, Siemens and Framatome have drawn together in designing and commercializing a common product initially destined for the export market which has become the EPR (European Pressurized water Reactor) project. The two companies have floated with equal participation the NPI (Nuclear Power International) sub-company to manage this project. The French and German utilities participate to the financing of the project, at present at the basic stage, and the safety authorities of both countries have carried out a joint evaluation of EPR safety. With a 1500 Mwe capacity, EPR will be equipped with advanced safety systems more performing than the existing systems. Conceivers want to maintain the economic competitiveness of EPR with respect to coal power plants. EPR will take over the oldest power plants by producing a safer and cheaper energy to provide for the needs of the developed countries and then of the developing countries with no risk for the environment. An enormous effort of communication must be carried out to reduce the public anxiety and to calm down the nuclear debate and show up its merits, in particular in Europe, where its contribution is vital. (J.S.)

  14. Effects of Language Background on Gaze Behavior: A Crosslinguistic Comparison Between Korean and German Speakers

    Science.gov (United States)

    Goller, Florian; Lee, Donghoon; Ansorge, Ulrich; Choi, Soonja

    2017-01-01

    Languages differ in how they categorize spatial relations: While German differentiates between containment (in) and support (auf) with distinct spatial words—(a) den Kuli IN die Kappe stecken (”put pen in cap”); (b) die Kappe AUF den Kuli stecken (”put cap on pen”)—Korean uses a single spatial word (kkita) collapsing (a) and (b) into one semantic category, particularly when the spatial enclosure is tight-fit. Korean uses a different word (i.e., netha) for loose-fits (e.g., apple in bowl). We tested whether these differences influence the attention of the speaker. In a crosslinguistic study, we compared native German speakers with native Korean speakers. Participants rated the similarity of two successive video clips of several scenes where two objects were joined or nested (either in a tight or loose manner). The rating data show that Korean speakers base their rating of similarity more on tight- versus loose-fit, whereas German speakers base their rating more on containment versus support (in vs. auf). Throughout the experiment, we also measured the participants’ eye movements. Korean speakers looked equally long at the moving Figure object and at the stationary Ground object, whereas German speakers were more biased to look at the Ground object. Additionally, Korean speakers also looked more at the region where the two objects touched than did German speakers. We discuss our data in the light of crosslinguistic semantics and the extent of their influence on spatial cognition and perception. PMID:29362644

  15. The current German regime governing third-party access to power transmission systems and denial of TPA, discussed from the angle of applicable civil law, energy industry law and antitrust law

    International Nuclear Information System (INIS)

    Kuehne, G.

    2000-01-01

    The German EnWG (energy industry law) for deregulation of the energy sector and implementation of the Internal Energy Market Directive of the EU contains an obligation to contract and make rules for establishing a legally binding system for access to and use of third parties of transmission and distribution networks in the competitive electricity market. The design of such contracts under private law as well as the grid code for network operation primarily being a matter of the contracting parties, the legal basis and opportunities for governmental supervisory functions are embodied in various laws. The legal analysis of this contribution examines the current situation and asks whether the existing provisions of the German BGB (Civil Code), antitrust law and the EnWG offer practicable means in case of need for governmental supervisory action in order to ensure evolution and adherence to a legal framework that will ensure the objectives of the politically willed deregulation of the energy sector and foster development of an open market serving the public welfare. (CB) [de

  16. Specific model for a gas distribution analysis in the containment at Almaraz NPP using GOTHIC computer code; Modelo de un sistema de control de gases combustibles en el edificio de contención, previsto específicamente para C.N. Almaraz con el código GOTHI

    Energy Technology Data Exchange (ETDEWEB)

    García González, M.; García Jiménez, P.; Martínez Domínguez, F.

    2016-07-01

    To carry out an analysis of the distribution of gases within the containment building at the CN Almaraz site, a simulation model with the thermohydraulic GOTHIC [1] code has been used. This has been assessed with a gas control system based on passive autocatalytic recombiners (PARs). The model is used to test the effectiveness of the control systems for gases to be used in the Almaraz Nuclear Power Plant, Uits I&II (Caceres, Spain, 1,035 MW and 1,044 MW). The model must confirm the location and number of the recombiners proposed to be installed. It is an essential function of the gas control system to avoid any formation of explosive atmospheres by reducing and limiting the concentration of combustible gases during an accident, thus maintaining the integrity of the containment. The model considers severe accident scenarios with specific conditions that produce the most onerous generation of combustible gases.

  17. Structural integrity assessment of a pressure container component. Design and service code implementation. Case studies; Evaluacion de la integridad estructural de un componente contenedor de presion. Aplicacion de los codigos para el disenio y servicio. Estudio de casos

    Energy Technology Data Exchange (ETDEWEB)

    Sanzi, H C [Universidad Tecnologica Nacional, Buenos Aires (Argentina)

    2006-07-01

    In the present work, the most important results of the local stresses occurred in the cracked pipes with a axial through-wall crack (outer), produced during operation of a Petrochemical Plant, using finite elements method, are presented. As requested, the component has been verified based 3D FE plastic analysis, under the postulated failure loading, assuring with this method a high degree of accuracy in the results. Codes used by Design and Service, as ASME Section VIII Div. 2 and API 579, have been used in the analysis. (author) [Spanish] La correcta evaluacion de la integridad estructural de componentes contenedores de presion y canierias requiere del conocimiento y la participacion de especialistas en 'Stress Analysis' y materiales e inspectores. En la actualidad, las tecnicas avanzadas de analisis, que incluyen un detallado 'Stress Analysis', a partir de la utilizacion del metodo de elementos finitos y la mecanica de Fractura, junto con el conocimiento del comportamiento de los materiales y la capacidad para detectar fisuras o discontinuidades - tales como los ensayos no destructivos y la emision acustica - permiten garantizar la seguridad de los componentes a lo largo de su vida util. En este camino los codigos de aplicacion, tanto en el disenio como en el servicio, son utilizados para llevar a cabo un estudio de integridad. En este trabajo se presenta un procedimiento de calculo para evaluar la integridad estructural de un componte contenedor de presion que posee una falla superficial no pasante, en donde se aplica el Codigo API 579, utilizando el metodo de elementos finitos y la mecanica de fractura. (autor)

  18. NPP Krsko containment environmental conditions during postulated accident

    International Nuclear Information System (INIS)

    Kozaric, M.; Cavlina, N.; Spalj, S.

    1989-01-01

    This paper presents NPP Krsko containment pressure and temperature increase during Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). Containment environmental condition calculation was performed by CONTEMPT4/MOD4 computer code. Design accident calculations were performed by RELAP4/MOD6 and RELAP5/MOD1 computer codes. Calculational abilities and application methodology of these codes are presented. The CONTEMPT code is described in more detail. The containment pressure and temperature time distribution are presented as well. (author)

  19. On the intonation of German intonation questions

    DEFF Research Database (Denmark)

    Petrone, Caterina; Niebuhr, Oliver

    2014-01-01

    German questions and statements are distinguished not only by lexical and syntactic but also by intonational means. This study revisits, for Northern Standard German, how questions are signalled intonationally in utterances that have neither lexical nor syntactic cues. Starting from natural......, but represents a separate attitudinal meaning dimension. Moreover, the findings support that both prenuclear and nuclear fundamental frequency (F0) patterns must be taken into account in the analysis of tune meaning....

  20. The Great War and German Memory

    DEFF Research Database (Denmark)

    Leese, Peter

    2012-01-01

    Review essay on Jason Crouthamel, The Great War and German Memory. Society, Politics and Psychological Trauma, 1914-18 (2009) and Anton Kaes, Shell Shock Cinema: Weimar Culture and the Wounds of War (2009)......Review essay on Jason Crouthamel, The Great War and German Memory. Society, Politics and Psychological Trauma, 1914-18 (2009) and Anton Kaes, Shell Shock Cinema: Weimar Culture and the Wounds of War (2009)...