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

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)

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

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.

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

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.

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

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

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)

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)

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

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.

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.

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

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,…

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.

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

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

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.

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.

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.)

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.

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

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…

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.)

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

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.)

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

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

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

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

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.)

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)

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.

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.

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

[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.

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

[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.

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)

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

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

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.

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

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)

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

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.

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

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.)

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)

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).

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.)

[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."…

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

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.

[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.

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)

[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.

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

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

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)

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.

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)

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)

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)

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

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

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...

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.)

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)

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

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.

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)

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.

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.)

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

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

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.)

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

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

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

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).

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)

[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.

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.

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.

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)

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)

[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.

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

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

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

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.)

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)

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?

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)

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

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.).

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

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)

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)

[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.

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)

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

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

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.

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)

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.

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

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)

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

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

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

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

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.

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)

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)

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

[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.

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

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

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

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)

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)

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.

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

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.

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

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.)

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

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)

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…

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

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

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

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

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.

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

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.)

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....

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

On the dimensioning of steel construction considering revision of German standards

International Nuclear Information System (INIS)

Lindhorst, L.; Schaudinn, L.; Milleder, J.

2005-01-01

Different procedures for dimensioning of structural steelwork are investigated in this paper. In the non-nuclear area, there has been made a change in the design concept for structural steelwork of surface constructions in Germany and the deterministic design concept of German DIN 18800 (03/81) [1] has been replaced by the semi-probabilistic design concept of DIN 18800 (11/90) [2]. Design codes established in nuclear safety standards for steelwork in German nuclear facilities are considered in the paper. Recent revisions of German nuclear safety standards and specifications are discussed with respect to regulations concerning design codes. The change in design concept for structural steelwork (non-nuclear) has not been adopted in recent revisions of nuclear safety standards and specifications for steelwork of components with importance to safety. The deterministic concept of DIN 18800 (03/81) [1] is compared with the semi-probabilistic concept of DIN 18800 (11/90) [2]. Differences between the standards [1] and [2] are investigated using an example of a steel construction. Results for dimensioning according to DIN 18800 (03/81) [1] are compared to those for dimensioning according to DIN 18800 (11/90) [2] putting the main emphasis on stress analysis and building structure interaction loads. (authors)

… 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…

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.

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.

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…

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

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…

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

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

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.

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

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.

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)

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.

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.)

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

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

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.

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.

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

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.

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

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

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

Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

International Nuclear Information System (INIS)

Kim, Tae H.; Yang, Chae Y.

2013-01-01

The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

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. 

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.

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

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...

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.

[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

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

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

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

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.

Shaded seafloor relief, backscatter strength, and surficial geology; German Bank, Scotian Shelf, offshore Nova Scotia

Science.gov (United States)

Todd, B.J.; Valentine, Page C.

2010-01-01

This map is part of a three-map series of German Bank, located on the Scotian Shelf off southern Nova Scotia.  This map is the product of a number of surveys (1997-2003) that used a multibeam sonar system to map 5321 km2 of the seafloor.  Other surveys collected geological data for scientific interpretation.  This map sheet shows the seafloor topography of German Bank in shaded-relief view and seafloor depth (coded by colour) at a scale of 1:1000,000.  Topographic contours generated from the multibeam data are shown (in white) on the colour-coded multibeam topography at a depth interval of 20 m.  Bathymetic contours (in blue) outside the multibeam survey area, presented at a depth interval of 10 m, are from the Natural Resource Map series (Canadian Hydrographic Service, 1967, 1971a, 1971b, 1972). Sheet 2 shows coloured backscatter strength in shaded-relief view.  Sheet 3 shows seafloor topography in shaded-relief view with colour-coded surficial geological units.

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…

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

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)

The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL

International Nuclear Information System (INIS)

Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.

1983-03-01

This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)

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.

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)

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

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.

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...

No grammatical gender effect on affective ratings: evidence from Italian and German languages.

Science.gov (United States)

Montefinese, Maria; Ambrosini, Ettore; Roivainen, Eka

2018-06-06

In this study, we tested the linguistic relativity hypothesis by studying the effect of grammatical gender (feminine vs. masculine) on affective judgments of conceptual representation in Italian and German. In particular, we examined the within- and cross-language grammatical gender effect and its interaction with participants' demographic characteristics (such as, the raters' age and sex) on semantic differential scales (affective ratings of valence, arousal and dominance) in Italian and German speakers. We selected the stimuli and the relative affective measures from Italian and German adaptations of the ANEW (Affective Norms for English Words). Bayesian and frequentist analyses yielded evidence for the absence of within- and cross-languages effects of grammatical gender and sex- and age-dependent interactions. These results suggest that grammatical gender does not affect judgments of affective features of semantic representation in Italian and German speakers, since an overt coding of word grammar is not required. Although further research is recommended to refine the impact of the grammatical gender on properties of semantic representation, these results have implications for any strong view of the linguistic relativity hypothesis.

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)

On-line monitoring and inservice inspection in codes

International Nuclear Information System (INIS)

Bartonicek, J.; Zaiss, W.; Bath, H.R.

1999-01-01

The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [de

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.

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

[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.

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.

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

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.)

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.

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

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)

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

[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

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....

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.

[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.

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.)

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.)

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

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

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

Chemistry of water and steam in power plants and related technologies. Glossary of terms and definitions English - German; German - English

Energy Technology Data Exchange (ETDEWEB)

Schmitz, H.P.; Teutenberg, U.

2006-07-01

This new edition of a technical dictionary is an evaluation of the technical terms found in the domestic and foreign literature and in information brochures of specialist firms, directives, guidelines, standards, etc. This dictionary contains more than 3,000 terms mainly with definitions with respect to the chemistry of water and steam in power plants along with the related types of water (untreated water, feedwater and boiler water, make-up water, waste water) and the water treatment processes (ion exchange, membrane process, etc.), water conditioning and chemical analysis, internal cleaning of steam generating plants (e.g. flushing, boiling-out, pre-operational and operational acid cleaning, steam blowing) as well as fundamentals of water chemistry. The technical knowledge of the authors, Heinz-Peter Schmitz, FDBR, with more than 25 years professional experience as translator/official in charge of documentation and Ulrich Teutenberg, Babcock/Hitachi with more than 30 years professional experience as senior consultant for water chemistry and commissioning is reflected in this dictionary. Part 1 contains the English-German version, Part 2 the German-English version. (orig.)

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…

[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.

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

"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...

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.)

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.

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

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

Mutation analysis of the MDM4 gene in German breast cancer patients

International Nuclear Information System (INIS)

Reincke, Scarlett; Govbakh, Lina; Wilhelm, Bettina; Jin, Haiyan; Bogdanova, Natalia; Bremer, Michael; Karstens, Johann H; Dörk, Thilo

2008-01-01

MDM4 is a negative regulator of p53 and cooperates with MDM2 in the cellular response to DNA damage. It is unknown, however, whether MDM4 gene alterations play some role in the inherited component of breast cancer susceptibility. We sequenced the whole MDM4 coding region and flanking untranslated regions in genomic DNA samples obtained from 40 German patients with familial breast cancer. Selected variants were subsequently screened by RFLP-based assays in an extended set of breast cancer cases and controls. Our resequencing study uncovered two MDM4 coding variants in 4/40 patients. Three patients carried a silent substitution at codon 74 that was linked with another rare variant in the 5'UTR. No association of this allele with breast cancer was found in a subsequent screening of 133 patients with bilateral breast cancer and 136 controls. The fourth patient was heterozygous for the missense substitution D153G which is located in a less conserved region of the MDM4 protein but may affect a predicted phosphorylation site. The D153G substitution only partially segregated with breast cancer in the family and was not identified on additional 680 chromosomes screened. This study did not reveal clearly pathogenic mutations although it uncovered two new unclassified variants at a low frequency. We conclude that there is no evidence for a major role of MDM4 coding variants in the inherited susceptibility towards breast cancer in German patients

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

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, ...

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.

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...

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

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)

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.

Comparison of radioactive doses after the last protection layer insight the reactor structure for Russian VVER-1000 and German PWR-1300 reactors

International Nuclear Information System (INIS)

Rahimi, A.; Mansourshaiflu, N.; Alizadeh, M. R.

2004-01-01

In pressurized reactors (VVER and PWR), various protections layers are used for reducing the output core doses. At any protection layer, some amount of neutron and gamma doses is reduced. In this project the axial flux of neutron and gamma beams have been evaluated at various protection layers in the operation state the German PWR-1300 and Russian VVER-1000 reactors by the MCNP computer code. For the purpose of effective use of the MCNP code and assuring its correct performance about of fluxed beams common and series of scientific answers and bench marks should be considered and the results obtained by the MCNP code, be compared with this answers. Then by using appropriate method, for reducing the flux variants of neutron and gamma beams at various protection layers of German PWR-1300 and Russian VVER-1000 reactors of the operation state of both reactors have been accelerated. In this projects, bench marks are computations and numbers existing in PSAR's present at Bushehr nuclear power plant. At the end, by using the results obtained and the standard doses, the time which a person can have work activity at the reactor wall (after the last protection layer), was compared for the operation status of the German PWR-1300 and Russian VVER-1000 reactors

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.)

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

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

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.

Loads on EPR containment after RPV failure at high pressure; Belastungen des EPR-Containments in Falle eines RDB-Versagens bei hohem Druck

Energy Technology Data Exchange (ETDEWEB)

Jacobs, G.

1995-08-01

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)

Release of gases and their influence on containment integrity during a hypothetical meltdown accident

International Nuclear Information System (INIS)

Hassmann, K.; Reimann, M.

1981-01-01

The sequence of a hypothetical core melt down accident has been subdivided into four phases. Heating up of the core until failure of the core support structure is the first phase. It starts at a certain water level in the reactor pressure vessel (RPV) and ends with the failure of the grid plate. The second phase is characterized by the evaporation of the water in the lower plenum of the RPV. The second phase lasts until a molten core debris is formed. The third phase is concerned with the heating up of the pressure vessel after formation of a molten pool in the lower plenum of the RPV. After pressure vessel failure, the molten corium will interact in the fourth phase with the concrete structure beneath the pressure vessel. In this paper the gas release during all four accident phases and the resulting pressure-time history within the containment of a German standard PWR is given, taking into account violent combustion of hydrogen. In particular, the differences caused by dsestruction of concrete with silicious and with calcareous aggregates has been analyzed. The basis for the results in the 4th phase is the WECHSL code. Long term containment calculations have been performed with the COCMEL-code

Explanatory models of addictive behaviour among native German, Russian-German, and Turkish youth.

Science.gov (United States)

Penka, S; Heimann, H; Heinz, A; Schouler-Ocak, M

2008-01-01

In Germany, the public system of addiction treatment is used less by migrants with addictive disorders than by their non-migrant counterparts. To date, the literature has focused primarily on language, sociocultural factors, and residence status when discussing access barriers to this part of the health care system. However, little attention has been paid to cultural differences in explanatory models of addictive behaviour. This is surprising when we consider the important role played by popular knowledge in a population's perceptions of and responses to illnesses, including their causes, symptoms, and treatment. In the present study, we examined explanatory models of addictive behaviour and of mental disorders in 124 native German und Russian-German youth and compared these models to those observed in an earlier study of 144 German and Turkish youth. We employed the free listing technique German and to compile the terms that participating subjects used to describe addictive behaviour. Subsequently, we examined how a subset of our study population assigned these terms to the respective disorders by means of the pile sort method. Although the explanatory models used by the German and Russian-German youth in our study were surprisingly similar, those employed by Turkish youth did not make any fundamental distinction between illegal and legal drugs (e.g. alcohol and nicotine). German and Russian-German youth regarded eating disorders as "embarrassing" or "disgraceful", but Turkish youth did not. Unlike our German and Russian-German subjects, the Turkish youth did not classify eating disorders as being addictive in nature. Moreover, medical concepts crucial to a proper understanding of dependence disorders (e.g. the term "physical dependence") were characterised by almost half of our Turkish subjects as useless in describing addictions. These findings show that it is impossible to translate medical or everyday concepts of disease and treatment properly into a different

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.

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.

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

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.)

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)

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R+D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR 1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Meseth, J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R and D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

International Nuclear Information System (INIS)

Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

1978-12-01

A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

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.

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, ...

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.

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

Pocket dictionary of laboratory equipment. English/German. Taschenwoerterbuch Laborausruestung. Deutsch/Englisch

Energy Technology Data Exchange (ETDEWEB)

Junge, H D

1987-01-01

This pocket dictionary contains the 2500 most common terms for scientific and technical equipment in chemical laboratories. It is a useful tool for those who are used to communicating in German and English, but have to learn the special terminology in this field.

DDR-Landeskundliches im Deutschunterricht (East German Culture in the German Lesson).

Science.gov (United States)

Wazel, Gerhard

1988-01-01

Demonstrates with concrete examples from the texts of a German class for advanced learners how aspects of East German culture can be realized in vocabulary, in lectures, and in the development of listening comprehension and speaking skills. Texts are appended. (DJD)

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)

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

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…

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)

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.

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

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.

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.)

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.)

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.)

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.)

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.

[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

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

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.

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.

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

Sequence Coding and Search System for licensee event reports: code listings. Volume 2

International Nuclear Information System (INIS)

Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

1985-04-01

Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. Volume 2 contains all valid and acceptable codes used for searching and encoding the LER data. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 2

Validation and application of the system code ATHLET-CD for BWR severe accident analyses

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

2016-10-15

Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

[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.

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

Accent, Intelligibility, and the Role of the Listener: Perceptions of English-Accented German by Native German Speakers

Science.gov (United States)

Hayes-Harb, Rachel; Watzinger-Tharp, Johanna

2012-01-01

We explore the relationship between accentedness and intelligibility, and investigate how listeners' beliefs about nonnative speech interact with their accentedness and intelligibility judgments. Native German speakers and native English learners of German produced German sentences, which were presented to 12 native German speakers in accentedness…

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...

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...

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".

Crystal chemistry of germanates: Characteristic structural features of Li,Ge-germanates

International Nuclear Information System (INIS)

Ilyushin, G.D.; Dem'yanets, L.N.

2000-01-01

Crystallochemical classification of eleven compounds from the Li-germanate family is suggested. Depending on the set of the primary building units (PBU) (M-octahedra of the composition [GeO 6 ] and T-tetrahedra of the composition [GeO 4 ]) and the type of their 'condensation', these germanates are divided into three crystallochemical groups: framework MT-structures (four phases), condensed MT-structures (two phases), and tetrahedral T-condensed structures (five phases). The structural characteristics of the framework Li,Ge-germanates are considered, i.e., their symmetry, crystallographically independent sets of the primary building units, framework architecture, and the types of chains and layers of the (Ge,O)-radicals

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)

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

International Nuclear Information System (INIS)

Glaeser, H.

1995-01-01

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

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)

German Command’s Guidelines on Training of Wehrmacht’s Military Units at Soviet-German Front Based on the Front-line Experience in 1941-1942

Directory of Open Access Journals (Sweden)

Shendrikov Evgeniy Aleksandrovich

2014-09-01

Full Text Available he article analyzes the guidelines of the training division of the General Staff of German land forces on improving the fighting techniques of Wehrmacht’s military units on the basis of the front-line experience gained in armed hostilities with the Red Army in 1941-1942. The study of archives allowed the author to reveal drawbacks and shortcomings of Wehrmacht’s military units training in the period preceding the second major German advance in Summer of 1942. In particular, the following drawbacks were revealed: incatious movement through monitored terrains, insufficient camouflage security, lack of proper skills in overhead shooting, poor field engineering of the infantry, poor land navigation, defeat of tank attacks and finally, tank combating at close range. The article also contains the characteristics of profound and thorough training of German command for the summer campaign of 1942, which covered all major issues such as the actions of reconnaissance forces, the organization of officers and non-commissioned officers’ training, close combat instructions, training of infantry’s field engineering, night training, making reports etc. At the end of the article the author comes to the conclusion that despite the intensified training and correction of previous mistakes, German command failed to achieve radical turning point in the war, which ended in the total defeat of fascist Germany.

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)

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

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)

FLUENT calculations of the hydrogen distribution in a containment during the OECD-NEA THAI HM-2 experiment

International Nuclear Information System (INIS)

Visser, D.C.; Komen, E.M.J.; Houkema, M.; Siccama, N.B.; Kyttaelae, Juha; Huhtanen, Risto; Takasuo, Eveliina

2009-01-01

Hydrogen may be released into the containment atmosphere of a nuclear power plant during a severe accident. Locally, high hydrogen concentrations may be reached that can possibly cause fast deflagration or even detonation and put the integrity of the containment at risk. Therefore, the distribution and mixing of hydrogen is an important safety issue for nuclear power plants. Computer codes can be applied to predict the hydrogen distribution in the containment within the course of a hypothetical severe accident and get an estimate of the local hydrogen concentration in the various zones of the containment. In this way the risk associated with the hydrogen safety issue can be determined, and safety related measurements and procedures could be assessed. In order to validate the existing computer codes in the context of hydrogen distribution in the containment of a nuclear power plant, experimental benchmark studies have been performed in the German Thermal-hydraulics, Hydrogen, Aerosols and Iodine (THAI) facility in the framework of the OECD-NEA THAI project. In order to demonstrate the capabilities of the commercial Computational Fluid Dynamics (CFD) code FLUENT the THAI HM-2 test was simulated independently by NRG and VTT. In the first phase of the HM-2 test a stratified hydrogen rich light gas layer was established in the upper part of the THAI containment. In the second phase steam was injected at a lower position inducing a rising plume that gradually dissolved the stratified hydrogen-rich layer from below. Thermo-dynamic phenomena like natural convection, mixing, condensation, heat transfer and distribution in different zones that are expected in severe accidents are involved. The calculated results by NRG and VTT (on hydrogen concentration, temperature, pressure and flow velocity) are compared to the experimental results. The most important differences between the CFD model of NRG and VTT are the computational mesh, condensation model and treatment of the solid

Mesh-based parallel code coupling interface

Energy Technology Data Exchange (ETDEWEB)

Wolf, K.; Steckel, B. (eds.) [GMD - Forschungszentrum Informationstechnik GmbH, St. Augustin (DE). Inst. fuer Algorithmen und Wissenschaftliches Rechnen (SCAI)

2001-04-01

MpCCI (mesh-based parallel code coupling interface) is an interface for multidisciplinary simulations. It provides industrial end-users as well as commercial code-owners with the facility to combine different simulation tools in one environment. Thereby new solutions for multidisciplinary problems will be created. This opens new application dimensions for existent simulation tools. This Book of Abstracts gives a short overview about ongoing activities in industry and research - all presented at the 2{sup nd} MpCCI User Forum in February 2001 at GMD Sankt Augustin. (orig.) [German] MpCCI (mesh-based parallel code coupling interface) definiert eine Schnittstelle fuer multidisziplinaere Simulationsanwendungen. Sowohl industriellen Anwender als auch kommerziellen Softwarehersteller wird mit MpCCI die Moeglichkeit gegeben, Simulationswerkzeuge unterschiedlicher Disziplinen miteinander zu koppeln. Dadurch entstehen neue Loesungen fuer multidisziplinaere Problemstellungen und fuer etablierte Simulationswerkzeuge ergeben sich neue Anwendungsfelder. Dieses Book of Abstracts bietet einen Ueberblick ueber zur Zeit laufende Arbeiten in der Industrie und in der Forschung, praesentiert auf dem 2{sup nd} MpCCI User Forum im Februar 2001 an der GMD Sankt Augustin. (orig.)

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.)

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

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.

How do German bilingual schoolchildren process German prepositions? - A study on language-motor interactions.

Science.gov (United States)

Ahlberg, Daniela Katharina; Bischoff, Heike; Strozyk, Jessica Vanessa; Bryant, Doreen; Kaup, Barbara

2018-01-01

While much support is found for embodied language processing in a first language (L1), evidence for embodiment in second language (L2) processing is rather sparse. In a recent study, we found support for L2 embodiment, but also an influence of L1 on L2 processing in adult learners. In the present study, we compared bilingual schoolchildren who speak German as one of their languages with monolingual German schoolchildren. We presented the German prepositions auf (on), über (above), and unter (under) in a Stroop-like task. Upward or downward responses were made depending on the font colour, resulting in compatible and incompatible trials. We found compatibility effects for all children, but in contrast to the adult sample, there were no processing differences between the children depending on the nature of their other language, suggesting that the processing of German prepositions of bilingual children is embodied in a similar way as in monolingual German children.

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...

A Comprehensive Approach to Post-Conflict Reconstruction: German Experience in Afghanistan

Directory of Open Access Journals (Sweden)

Philip Olegovich Trunov

2016-12-01

Full Text Available As a part of a larger research project on complex linkages between security and development implemented by the Center for Security and Development Studies at the Lomonosov Moscow State University, this paper examines the complex dilemmas of pursuing comprehensive approach to post-conflict reconstruction with an example of Federal Republic of Germany’s experience with the Provincial Reconstruction Teams (PRTs in Afghanistan after the terrorist attacks of 9/11. The first section depicts the background of deployment of the PRTs on Afghan territory. The second section compares the U.S., British and German models of PRTs and assesses the role of those teams as vehicles of close civil-military, interagency coordination. The third and the fourth sections identify respectively characteristic features of the German PRTs’ activities in Kunduz and Badakhshan provinces and the main obstacles that hindered the achievement of key objectives. The conclusion contains a concise assessment of effectiveness and efficiency of PRTs as a mechanism of addressing complex challenges of post-conflict reconstruction and transition to peaceful development. It also postulates that the PRT model, regardless of some objective difficulties faced by various German agencies, may be very relevant in the future in fulfilling a revised agenda of the German foreign and defense policies in the XXI century.

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

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

www.kernenergie.de - nuclear power has a German Internet address

International Nuclear Information System (INIS)

Anon.

2000-01-01

www.kernenergie.de is the address on the worldwide web under which the German nuclear organizations, Deutsches Atomforum (DAtF), Informationskreis Kernenergie (IK), and Kerntechnische Gesellschaft (KTG) as well as atw - internationale Zeitschrift fuer Kernenergie, and INFORUM can be reached. Extensive sources of information, discussions, on-line dictionaries, computer codes, dynamic web pages, digital documents and multimedia offerings can be called up via the portal under the individual web sites. In this way, www.kernenergie.de provides a comprehensive and up-to-date background of information about nuclear power and adjacent topics in the digital worldwide web. (orig.) [de

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)

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.)

Compliance with technical codes becomes obligatory for receipt of feed-in tariff and ancillary services bonus for wind power plants in Germany

Energy Technology Data Exchange (ETDEWEB)

Boemer, Jens. C. [Ecofys Germany GmbH, Berlin (Germany). Power Systems and Markets Group; Delft Univ. of Technology (Netherlands). Electrical Power Systems Group; Burges, Karsten [Ecofys Germany GmbH, Berlin (Germany). Power Systems and Markets Group; Kumm, Thomas [VDE, Network Technology and Operation Forum (FNN), Berlin (Germany); Poeller, Markus [DIgSILENT GmbH, Gomaringen (Germany)

2009-07-01

The German government aims at a share of electricity generation based on renewable sources (RES-E) of more than 30% by 2020. A major part of this share will be reached by the use of wind power plants (WPP). Therefore, when amending the Renewable Energy Sources Act (EEG) in June 2008, the German government acknowledged the importance of enhanced technical requirements for WPP and announced the development of a respective technical ordinance. For all WPP that go into operation after June 2010, full grid code-compliance becomes a precondition for privileged network access, receipt of feed-in tariff, and extra payments (''ancillary services bonus''). The basis for the technical requirements in the EEG has been laid down by Medium-Voltage (MV)-Directive 2008 and the TransmissionCode 2007. However, the work of the authors - consulting the German government during the development of the technical ordinance - showed that the TransmissionCode 2007 needed careful review and some clarifying specifications were proposed. Eventually, the so called ''Ancillary Services Ordinance'' for wind power plants went into power in July 2009. The interrelation between all three grid codes as well as their specifications and implications for renewable energy sources generators, with special regard to wind power plants, are presented in this paper. (orig.)

[Sociotherapy in German social law. Indication, contents, and aspects of public health].

Science.gov (United States)

Frieboes, R-M

2003-07-01

In German mental health services, the ill-defined term "sociotherapy" has been used to designate nonmedical, social, and work-related components of the care process. Recently, a new component of outpatient/community mental health care called "sociotherapy" (according to Paragraph 37a of the Fifth German Social Code) which is funded by the public health insurance system has been introduced and is now in the process of being implemented. The paper describes (a) patients eligible for the service and (b) the aims and scope of this case management module. The key objectives are to motivate patients with schizophrenia to utilise mental health services and antipsychotic medication and to liaise with psychosocial services. Therefore, sociotherapy is distinct from (a) multidisciplinary inpatient care for people with severe mental illness, (b) assertive community treatment, (c) community care provided by social workers or community psychiatric nurses, and (d) family interventions. So far there has been little evaluation of sociotherapy.

The German Radiation Protection Ordinance (StrlSchV). 3. ed.

International Nuclear Information System (INIS)

Hinrichs, O.

1992-01-01

The German Radiation Protection Ordinance constitutes the central statutory instrument containing the main protective provisions for all fields of application of radioactive materials and ionizing radiation, with the exception of the field covered by the X-ray Ordinance. The current text of the Ordinance is that promulgated on 30.06.1989 with the subsequent amendments, as last amended through the Unification Treaty (Einigungsvertrag) of 23.09.1990. The Radiation Protection Ordinance was adopted on the basis of the German Nuclear Energy Act (Atomgesetz), which contains, inter alia, the necessary empowerments to issue statutory ordinances. Further fields containing relevant protective provisions are, above all, the law of dangerous substances (which concerns the transport of radioactive materials), the law of pharmaceutical products and the law of foodstuffs. The whole regulatory package is a part of EC and Euratom law. The limit values of the Euratom Directives were transposed into the Radiation Protection Ordinance. In order to reduce the bulk of the Ordinance, the legislator has only included the limit values for the most important radionuclides, and has made provision for the separate promulgation of the other limit values. In order to enhance the practical use of the book, the provisions governing the radiation pass, which are regulated in a separate administrative instruction, are also reproduced. (orig./HSCH) [de

Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2004-07-01

MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

International Nuclear Information System (INIS)

Lee, Young Jin; Chung, Bub Dong

2004-01-01

MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

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)

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...

How do German bilingual schoolchildren process German prepositions? – A study on language-motor interactions

Science.gov (United States)

Bischoff, Heike; Strozyk, Jessica Vanessa; Bryant, Doreen; Kaup, Barbara

2018-01-01

While much support is found for embodied language processing in a first language (L1), evidence for embodiment in second language (L2) processing is rather sparse. In a recent study, we found support for L2 embodiment, but also an influence of L1 on L2 processing in adult learners. In the present study, we compared bilingual schoolchildren who speak German as one of their languages with monolingual German schoolchildren. We presented the German prepositions auf (on), über (above), and unter (under) in a Stroop-like task. Upward or downward responses were made depending on the font colour, resulting in compatible and incompatible trials. We found compatibility effects for all children, but in contrast to the adult sample, there were no processing differences between the children depending on the nature of their other language, suggesting that the processing of German prepositions of bilingual children is embodied in a similar way as in monolingual German children. PMID:29538404

The opening of the debates on the extension of German nuclear plant lifetime

International Nuclear Information System (INIS)

Guidez, J.

2010-01-01

After having recalled that the German nuclear plants were to be progressively stopped by 2021, the author reports that the political coalition has thrown this Agenda back into question, because the development of alternative energies is insufficient to compensate the programmed closing down, and the German government wants to avoid the use of thermal plants which are generating CO 2 emissions. Negotiations are underway between the German government and the four electric companies in charge of different plants. It appears that even plants which were to be stopped in 2010, could keep on operating. The report contains tables indicating, for each nuclear plant, its type, its operator, its power, its dates of beginning and programmed end of commercial activity, its production variation in time. Some comments are given about already stopped reactors (Brunsbuttel and Krummel) which faced some problems in the past, and about waste storing sites (Asse, Gorleben, Morsleben, Konrad)

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

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

Cross-linguistic vowel variation in trilingual speakers of Saterland Frisian, Low German, and High German.

Science.gov (United States)

Peters, Jörg; Heeringa, Wilbert J; Schoormann, Heike E

2017-08-01

The present study compares the acoustic realization of Saterland Frisian, Low German, and High German vowels by trilingual speakers in the Saterland. The Saterland is a rural municipality in northwestern Germany. It offers the unique opportunity to study trilingualism with languages that differ both by their vowel inventories and by external factors, such as their social status and the autonomy of their speech communities. The objective of the study was to examine whether the trilingual speakers differ in their acoustic realizations of vowel categories shared by the three languages and whether those differences can be interpreted as effects of either the differences in the vowel systems or of external factors. Monophthongs produced in a /hVt/ frame revealed that High German vowels show the most divergent realizations in terms of vowel duration and formant frequencies, whereas Saterland Frisian and Low German vowels show small differences. These findings suggest that vowels of different languages are likely to share the same phonological space when the speech communities largely overlap, as is the case with Saterland Frisian and Low German, but may resist convergence if at least one language is shared with a larger, monolingual speech community, as is the case with High German.

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

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.

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

Public Management Reform without Managers: The Case of German Public Schools

Science.gov (United States)

Mintrop, Rick

2015-01-01

Purpose: The purpose of this paper is to discuss the role of principals in light of public management reforms taking place in the German educational system and in reference to the empirical patterns uncovered by the papers contained in the Special Issue. Policy makers have created new expectations and new technologies that seem to suggest to…

Review on the present status of the German Risk Study: overall results

International Nuclear Information System (INIS)

Heuser, F.W.

1982-01-01

An overview on the most important results of the German Risk Study (Phase A) is given. Besides plant design analysis, the description mainly refers to core melt accidents, radioactive release and accidental consequences depending on various types of containment failure. Thereby some differences or results compared to WASH 1400 are discussed

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

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…

Interfacial and Wall Transport Models for SPACE-CAP Code

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun

2009-01-01

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code

Interfacial and Wall Transport Models for SPACE-CAP Code

Energy Technology Data Exchange (ETDEWEB)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2009-10-15

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

The participation of the German Länder in the EU decision-making process

Directory of Open Access Journals (Sweden)

Carlo Panara

2012-09-01

Full Text Available This article sketches out the main features of the German system of regional participation in the EU decision-making process with the aim of identifying the key strengths and weaknesses of that system. The author will be examining whether the model currently in place allows for a sufficiently rapid and effective response to the EU decision-making process, and whether there is a sufficiently balanced representation of regional and federal interests at the EU level. The article will also contain a proposal as to how the German model of regional participation in the EU decision-making process could be further improved.

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

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.

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

Code for plant identification (KKS) key in PC version

International Nuclear Information System (INIS)

Pannenbaecker, K.

1991-01-01

The plant identification system (KKS) as a common development of german plant operators, erection firms and also power plant oriented organisations have decisively influenced the technical-organizing activities of planning and erections as operations and maintenance of all kind of power plants. Fundamentals are three key parts, operation, armatures and function keys. Their management and application is executed by a plantidentification-key code in a PC version, which is briefly described in this report. (orig.) [de

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…

Prospects for research in haemophilia with real-world data-An analysis of German registry and secondary data.

Science.gov (United States)

Schopohl, D; Bidlingmaier, C; Herzig, D; Klamroth, R; Kurnik, K; Rublee, D; Schramm, W; Schwarzkopf, L; Berger, K

2018-02-28

Open questions in haemophilia, such as effectiveness of innovative therapies, clinical and patient-reported outcomes (PROs), epidemiology and cost, await answers. The aim was to identify data attributes required and investigate the availability, appropriateness and accessibility of real-world data (RWD) from German registries and secondary databases to answer the aforementioned questions. Systematic searches were conducted in BIOSIS, EMBASE and MEDLINE to identify non-commercial secondary healthcare databases and registries of patients with haemophilia (PWH). Inclusion of German patients, type of patients, data elements-stratified by use in epidemiology, safety, outcomes and health economics research-and accessibility were investigated by desk research. Screening of 676 hits, identification of four registries [national PWH (DHR), national/international paediatric (GEPARD, PEDNET), international safety monitoring (EUHASS)] and seven national secondary databases. Access was limited to participants in three registries and to employees in one secondary database. One registry asks for PROs. Limitations of secondary databases originate from the ICD-coding system (missing: severity of haemophilia, presence of inhibitory antibodies), data protection laws and need to monitor reliability. Rigorous observational analysis of German haemophilia RWD shows that there is potential to supplement current knowledge and begin to address selected policy goals. To improve the value of existing RWD, the following efforts are proposed: ethical, legal and methodological discussions on data linkage across different sources, formulation of transparent governance rules for data access, redefinition of the ICD-coding, standardized collection of outcome data and implementation of incentives for treatment centres to improve data collection. © 2018 John Wiley & Sons Ltd.

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...

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

Containment long-term operational integrity

International Nuclear Information System (INIS)

Sammataro, R.F.

1990-01-01

Periodic integrated leak rate tests are required to assure that containments continue to meet allowable leakage limits. Although overall performance has been quite good to date, several major containment aging and degradation mechanisms have been identified. Two pilot plant life extension (PLEX) studies serve as models for extending the operational integrity of present containments for light-water cooled nuclear power plants in the United States. One study is for a Boiling-Water Reactor (BWR) and the second is for a Pressurized-Water Reactor (PWR). Research and testing programs for determining the ultimate pressure capacity and failure mechanisms for containments under severe loading conditions and studies for extending the life of current plants beyond the present 40-year licensed lifetime are under way. This paper presents an overview of containment designs in the United States. Also presented are a discussion of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) and regulatory authority requirements for the design, construction, inservice inspection, leakage testing and repair of steel and concrete containments. Findings for containments from the pilot PLEX studies and continuing containment integrity research and testing programs are discussed. The ASME Code and regulatory requirements together with recommendations from the PLEX studies and containment integrity research and testing provide a basis for continued containment long-term operational integrity. (orig./GL)

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)

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.

A simple route to synthesize manganese germanate nanorods

International Nuclear Information System (INIS)

Pei, L.Z.; Yang, Y.; Yuan, C.Z.; Duan Taike; Zhang Qianfeng

2011-01-01

Manganese germanate nanorods have been synthesized by a simple route using germanium dioxide and manganese acetate as the source materials. X-ray diffraction observation shows that the nanorods are composed of orthorhombic and monoclinic manganese germanate phases. Scanning electron microscopy and transmission electron microscopy observations display that the manganese germanate nanorods have flat tips with the length of longer than 10 micrometers and diameter of 60-350 nm, respectively. The role of the growth conditions on the formation of the manganese germanate nanorods shows that the proper selection and combination of the growth conditions are the key factor for controlling the formation of the manganese germanate nanorods. The photoluminescence spectrum of the manganese germanate nanorods exhibits four fluorescence emission peaks centered at 422 nm, 472 nm, 487 nm and 530 nm showing the application potential for the optical devices. - Research Highlights: → Manganese germanate nanorods have been synthesized by simple hydrothermal process. → The formation of manganese germanate nanorods can be controlled by growth conditions. → Manganese germanate nanorods exhibit good PL emission ability for optical device.

Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Development of thermal hydraulic evaluation code for CANDU reactors

International Nuclear Information System (INIS)

Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

2004-02-01

To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

The first accident simulation of Angra-1 power plant using the ALMOD computer code

International Nuclear Information System (INIS)

Camargo, C.T.M.

1981-01-01

The implementation of the german computer code ALMOD and its application in the calculation of Angra-1, a nuclear power plant different from the KWU power plants, demanded study and models adaptation, and due to economic reasons simplifications and optimizations were necessary. The first results define the analytical potential of the computer code, confirm the adequacy of the adaptations done and provide relevant conclusions about the Angra-1 safety analysis, showing at the same time areas in which the model can be applied or simply improved. (E.G.) [pt

Joint crisis plans and psychiatric advance directives in German psychiatric practice.

Science.gov (United States)

Radenbach, Katrin; Falkai, Peter; Weber-Reich, Traudel; Simon, Alfred

2014-05-01

This study explores the attitude of German psychiatrists in leading positions towards joint crisis plans and psychiatric advance directives. This topic was examined by contacting 473 medical directors of German psychiatric hospitals and departments. They were asked to complete a questionnaire developed by us. That form contained questions about the incidence and acceptance of joint crisis plans and psychiatric advance directives and previous experiences with them. 108 medical directors of psychiatric hospitals and departments responded (response rate: 22.8%). Their answers demonstrate that in their hospitals these documents are rarely used. Among the respondents, joint crisis plans are more accepted than psychiatric advance directives. There is a certain uncertainty when dealing with these instruments. Our main conclusion is that German psychiatry needs an intensified discussion on the use of instruments for patients to constitute procedures for future critical psychiatric events. For this purpose it will be helpful to collect more empirical data. Furthermore, the proposal of joint crisis plans in psychiatric hospitals and departments should be discussed as well as the possibility of consulting an expert during the preparation of a psychiatric advance directive.

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

International Nuclear Information System (INIS)

Camargo, C.T.M.

1981-02-01

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

Sodium/water pool-deposit bed model of the CONACS code

International Nuclear Information System (INIS)

Peak, R.D.

1983-01-01

A new Pool-Bed model of the CONACS (Containment Analysis Code System) code represents a major advance over the pool models of other containment analysis code (NABE code of France, CEDAN code of Japan and CACECO and CONTAIN codes of the United States). This new model advances pool-bed modeling because of the number of significant materials and processes which are included with appropriate rigor. This CONACS pool-bed model maintains material balances for eight chemical species (C, H 2 O, Na, NaH, Na 2 O, Na 2 O 2 , Na 2 CO 3 and NaOH) that collect in the stationary liquid pool on the floor and in the desposit bed on the elevated shelf of the standard CONACS analysis cell

76 FR 46805 - Notice of Utah Adoption by Reference of the Pesticide Container Containment Rule

Science.gov (United States)

2011-08-03

... ENVIRONMENTAL PROTECTION AGENCY [FRL-9447-8] Notice of Utah Adoption by Reference of the Pesticide Container Containment Rule AGENCY: Environmental Protection Agency (EPA). ACTION: Notice. SUMMARY: This... Pesticide Container Containment (PCC) Rule regulations. In accordance with State of Utah Agricultural Code...

Conclusions Drawn from the Investigation of LOCA-Induced Insulation Debris Generation and its Impact on Emergency Core Cooling (ECC) at German NPPs - Approach Taken by / Perspective of The German TSO (TuV)

International Nuclear Information System (INIS)

Huber, J.

2004-01-01

Initiated by the Barsebaeck incident in 1992 and the following activities related to the LOCA-induced insulation debris generation and its impact on emergency core cooling, investigations on German PWRs and BWRs were performed in these areas. The investigations on the German BWRs were carried out in detail immediately after the Barsebaeck incident in the years 1992 through 1994. Detailed investigations on the German PWRs started after the issue of the OECD report in 1996. Therefore the investigations on the impact of LOCA-induced insulation debris generation on strainer plugging carried out in Germany in the last years were focused mainly on the German PWRs. In the framework of these investigations of the German PWRs, the relevant parameters and phenomena were investigated in detail by the plant owners in the years from 1997 through 1999. The results were summarised in reports for each plant. The main results of the investigations conducted by the plant owners were that the plant owners considered backfitting in German PWRs is not necessary to guarantee emergency core cooling following a LOCA with insulation debris generation. As the technical support organisation for the German Bavarian and Hessian state authority, the TUV Suddeutschland was called upon to examine these investigations and the conclusions drawn by the plant owners. We compared each of the parameters and phenomena against the state of knowledge. The results of our examination in 1999 showed that the investigations of the plant owners were generally correct, but we stated also, that due to existing uncertainties, further investigations are necessary to validate the results. To meet these demands, the plant owners installed a working group for planning and performing newer, more realistic large-scale experiments (scaling factor 1:4) to investigate the transport mechanism of the insulation material within the containment sump, the head loss at the strainers and to estimate the amount of insulation

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

International Nuclear Information System (INIS)

Sdouz, Gert; Mayrhofer, Robert

2009-01-01

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

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…

German risk study for nuclear power plants. Vol. 5

International Nuclear Information System (INIS)

Bracht, K.F.; Javeri, V.; Keusenhoff, J.; Meier, S.; Roehrs, W.; Scharfe, A.; Tiltmann, M.; Mayinger, F.

1980-01-01

This appendix contains the analyses on core melt accidents carried out for the German risk study. These accidents constitute a major portion of the event sequences relevant to the risk. It deals, in particular, with the processes involved in the reactor core melt, the behaviour of the containment vessel including the possible types of failures, and the consequences of possible steam explosions. The investigations were performed for limiting cases for which a complete failure of the emergency core cooling and residual heat removal system is anticipated. Under these conditions, the reactor core melts down and penetrates the reactor pressure vessel bottom. In the course of the following interaction between core melt and concrete foundation, the melt is assumed to come into contact with the sump water. The evaporation of the sump water leads to an overpressure failure of the containment vessel within approximately one day after accident initiation. A destruction of the containment as a consequence of a steam explosion is considered highly unprobable. (orig.) [de

Users guide for NRC145-2 accident assessment computer code

International Nuclear Information System (INIS)

Pendergast, M.M.

1982-08-01

An accident assessment computer code has been developed for use at the Savannah River Plant. This computer code is based upon NRC Regulatory Guide 1.145 which provides guidence for accident assessements for power reactors. The code contains many options so that the user may utilize the code for many different assessments. For example the code can be used for non-nuclear assessments such as Sulpher Dioxide which may be required by the EPA. A discription of the code is contained in DP-1646. This document is a compilation of step-by-step instructions on how to use the code on the SRP IBM 3308 computer. This document consists of a number of tables which contain copies of computer listings. Some of the computer listings are copies of input; other listings give examples of computer output

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.

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)

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.)

Containment integrity analysis under accidents

International Nuclear Information System (INIS)

Lin Chengge; Zhao Ruichang; Liu Zhitao

2010-01-01

Containment integrity analyses for current nuclear power plants (NPPs) mainly focus on the internal pressure caused by design basis accidents (DBAs). In addition to the analyses of containment pressure response caused by DBAs, the behavior of containment during severe accidents (SAs) are also evaluated for AP1000 NPP. Since the conservatism remains in the assumptions,boundary conditions and codes, margin of the results of containment integrity analyses may be overestimated. Along with the improvements of the knowledge to the phenomena and process of relevant accidents, the margin overrated can be appropriately reduced by using the best estimate codes combined with the uncertainty methods, which could be beneficial to the containment design and construction of large passive plants (LPP) in China. (authors)

Study of the sodium oxide aerosol behavior in a containment building

International Nuclear Information System (INIS)

Fermandjian, Jean.

1982-01-01

Fast breeder reactor safety analysis needs validated models to predict the behavior of sodium oxide aerosols (released during hypothetical accidents) in the different reactor compartments and, in particular, to evaluate the quantity (and the size distribution) of the particles which can be released outside the containment building by taking into account the associated devices (ventilation, especially). In order to validate the computer models, experimental tests were performed by CEA/DSN (Atomic Energy Commission/Nuclear Safety Department) at Cadarache: sodium pool fire tests in a 400 m 3 concrete vessel (CASSANDRE tests) and in a 4.4 m 3 steel vessel (EMIS tests). Furthermore, we utilized the results of the experimental tests performed by our German colleagues of KFK/LAF (Kernforschungszentrum Karlsruhe/Laboratorium fuer Aerosolphysik und Filtertechnik): sodium pool fire tests in a 220 m 3 steel vessel (FAUNA tests). The application of the computer models to the experimental tests and to the reactor case allowed us: - to explain the differences observed between the results given by the codes ''with log-normal hypothesis'' (HAARM 3) and ''with finite difference'' (PARDISEKO 3B), - to define the limits of application of the codes ''with log-normal hypothesis'', - to determine the important parameters (characteristics of the aerosol source, form factors of the agglomerates and collision efficiency) and the important phenomena (turbulent coagulation, in particular) [fr

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.

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

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.)

German congress on X-ray radiography 1981. Deutscher Roentgenkongress 1981

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

The book contains abstracts of 280 papers read at the 1981 German X-ray Congress - some abstracts, unfortunately, are not given. The following subjects were discussed: Diagnosis of internal injuries, traumatic lesions of the skull and brain, diagnosis in orthopaedics and ophthalmology, gastric examinations, diagnosis of pancreatitis due to alcoholism, ultrasonographic diagnosis, biopsy, CT diagnosis of the pelvis and nephro-urology, CT-diagnosis of the thorax, radiotherapy, nuclear medicine, radiobiology, radiological techniques and other subjects.

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

Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code

International Nuclear Information System (INIS)

Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.

2004-01-01

Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials

Evaluation of results of Phase B of the German Risk Study for Nuclear Power Plants

International Nuclear Information System (INIS)

Fischer, B.; Hahn, L.; Sailer, M.

1989-01-01

The expert opinion summarizes and discusses the intermediate and the final results of Phase B of the German Risk Study. Emphasis is placed on systems analyses and event scenarios, core melt-down scenarios and effects on the containment, failure of the containment and resulting source terms, radioactivity transport and consequences of accidents, multi-system emergencies and external disturbances, accident management and means of mitigation of damage. (DG) [de

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.

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

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

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)

CONTRANS 2 code conversion from Apollo to HP

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-01-01

CONTRANS2 computer code is used to calculate transient thermal hydraulic responses of containment building to loss of coolant and main steam line break accident. Mass and energy release to the containment following an accident are code inputs. This report firstly describes detailed work carried out for installation of CONTRANS2 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of CONTRANS2 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

The reliability of the German version of the Richards Campbell Sleep Questionnaire.

Science.gov (United States)

Krotsetis, Susanne; Richards, Kathy C; Behncke, Anja; Köpke, Sascha

2017-07-01

The assessment of sleep quality in critically ill patients is a relevant factor of high-quality care. Despite the fact that sleep disturbances and insufficient sleep management contain an increased risk of severe morbidity for these patients, a translated and applicable instrument to evaluate sleep is not available for German-speaking intensive care settings. This study aimed to translate the Richards Campbell Sleep Questionnaire (RCSQ), a simple and validated instrument eligible for measuring sleep quality in critically ill patients, and subsequently to evaluate the internal consistency of the German version of the RCSQ. Furthermore, it also aimed to inquire into the perception of sleep in a sample of critically ill patients. The RCSQ was translated following established methodological standards. Data were collected cross-sectionally in a sample of 51 patients at 3 intensive care units at a university hospital in Germany. The German version of the RCSQ showed an overall internal consistency (Cronbach's alpha) of 0·88. The mean of the RSCQ in the sample was 47·00 (SD ± 27·57). Depth of sleep was rated the lowest and falling asleep again the highest of the RCSQ sleep items. The study demonstrated very good internal consistency of the German version of the RCSQ, allowing for its application in practice and research in German-speaking countries. Quality of sleep perception was generally low in this sample, emphasizing the need for enhanced care concepts regarding the sleep management of critically ill patients. Relevance to clinical practice Assessment of self-perception of sleep is crucial in order to plan an individually tailored care process. © 2017 British Association of Critical Care Nurses.

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.

MELCOR/CONTAIN LMR Implementation Report-Progress FY15

Energy Technology Data Exchange (ETDEWEB)

Humphries, Larry L.; Louie, David

2016-01-01

This report describes the progress of the CONTAIN-LMR sodium physics and chemistry models to be implemented in to MELCOR 2.1. It also describes the progress to implement these models into CONT AIN 2 as well. In the past two years, the implementation included the addition of sodium equations of state and sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laborat ory by modifying MELCOR to include liquid lithium equation of state as a working fluid to mode l the nuclear fusion safety research. The second source uses properties generated for the SIMMER code. Testing and results from this implementation of sodium pr operties are given. In addition, the CONTAIN-LMR code was derived from an early version of C ONTAIN code. Many physical models that were developed sin ce this early version of CONTAIN are not captured by this early code version. Therefore, CONTAIN 2 is being updated with the sodium models in CONTAIN-LMR in or der to facilitate verification of these models with the MELCOR code. Although CONTAIN 2, which represents the latest development of CONTAIN, now contains ma ny of the sodium specific models, this work is not complete due to challenges from the lower cell architecture in CONTAIN 2, which is different from CONTAIN- LMR. This implementation should be completed in the coming year, while sodi um models from C ONTAIN-LMR are being integrated into MELCOR. For testing, CONTAIN decks have been developed for verification and validation use. In terms of implementing the sodium m odels into MELCOR, a separate sodium model branch was created for this document . Because of massive development in the main stream MELCOR 2.1 code and the require ment to merge the latest code version into this branch, the integration of the s odium models were re-directed to implement the sodium chemistry models first. This change led to delays of the actual implementation. For aid in the future implementation of sodium

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

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...

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

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

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)

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)

MELCOR/CONTAIN LMR Implementation Report. FY14 Progress

Energy Technology Data Exchange (ETDEWEB)

Humphries, Larry L; Louie, David L.Y.

2014-10-01

This report describes the preliminary implementation of the sodium thermophysical properties and the design documentation for the sodium models of CONTAIN-LMR to be implemented into MELCOR 2.1. In the past year, the implementation included two separate sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laboratory by modifying MELCOR to include liquid lithium equation of state as a working fluid to model the nuclear fusion safety research. To minimize the impact to MELCOR, the implementation of the fusion safety database (FSD) was done by utilizing the detection of the data input file as a way to invoking the FSD. The FSD methodology has been adapted currently for this work, but it may subject modification as the project continues. The second source uses properties generated for the SIMMER code. Preliminary testing and results from this implementation of sodium properties are given. In this year, the design document for the CONTAIN-LMR sodium models, such as the two condensable option, sodium spray fire, and sodium pool fire is being developed. This design document is intended to serve as a guide for the MELCOR implementation. In addition, CONTAIN-LMR code used was based on the earlier version of CONTAIN code. Many physical models that were developed since this early version of CONTAIN may not be captured by the code. Although CONTAIN 2, which represents the latest development of CONTAIN, contains some sodium specific models, which are not complete, the utilizing CONTAIN 2 with all sodium models implemented from CONTAIN-LMR as a comparison code for MELCOR should be done. This implementation should be completed in early next year, while sodium models from CONTAIN-LMR are being integrated into MELCOR. For testing, CONTAIN decks have been developed for verification and validation use.

2015 German refrigeration and air conditioning meeting. Abstracts; Deutsche Kaelte- und Klimatagung 2015. Kurzfassungen

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-07-01

The volume contains the abstracts of the 2015 German refrigeration and air conditioning meeting in 5 chapters: cryo-technology, fundamentals of materials for refrigeration engineering and heat pump technology, facilities and components for the refrigeration and heat pump technology; application of refrigeration engineering; air conditioning technology and heat pump application.

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.

Dictionary of pressure vessel and piping technology. English-German. Woerterbuch der Druckbehaelter- und Rohrleitungstechnik. Englisch-Deutsch

Energy Technology Data Exchange (ETDEWEB)

Jentgen, L; Schmitz, H P

1986-01-01

A specialised dictionary has been compiled containing the appropriate English and German terms in the following technical fields: materials science, welding, destructive and non-destructive testing, thermal and mass transfer, the design and construction in particular of pressure vessels, tanks, heat exchangers, piping, expansion joints, valves, and components associated with the above fields. This dictionary is the result of many years spent in evaluating technical terminology from the relevant American and British regulations, technical rules, standards, and specifications (see bibliography) and correlating these with the terminology of comparable German regulations, rules and standards, together with the essential technical literature.

Upgrades to the WIMS-ANL code

International Nuclear Information System (INIS)

Woodruff, W. L.

1998-01-01

The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries

Upgrades to the WIMS-ANL code

International Nuclear Information System (INIS)

Woodruff, W.L.; Leopando, L.S.

1998-01-01

The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries. (author)

The German model of capitalism and the persistence of outward foreign direct investment: evidence from German manufacturing industries

Directory of Open Access Journals (Sweden)

Martin T Bohl

2011-06-01

Full Text Available Against the backdrop of critique on the German model of capitalism in general, and German public policy in particular as to the ability to successfully adjust to rapid change and exogenous shocks in wake of economic globalisation, this paper investigates the degree of shock persistence in foreign direct investment (FDI of ten German manufacturing industries for the period 1976 to 2003. Theory on exports and non-FDI investment suggests that FDI should exhibit a considerable degree of shock persistence because they are subject to high sunk costs because of high entry and exit costs associated with the high level of asset specificity that is normally connected to FDI. Persistence in foreign direct investment time series data is established by applying various unit root tests. The results are robust to the potential presence of structural breaks in the data. The empirical analysis shows that German outward FDI in mature manufacturing industries, with one exception, exhibits a high degree of shock persistence. The results suggest, at least for mature German industries, that the sunk costs view on shock persistency is confirmed for outward FDI. The results furnish evidence for a tentative assessment of the relationship between German public policy and FDI strategies of multinational firms.

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)

Non-binary Entanglement-assisted Stabilizer Quantum Codes

OpenAIRE

Riguang, Leng; Zhi, Ma

2011-01-01

In this paper, we show how to construct non-binary entanglement-assisted stabilizer quantum codes by using pre-shared entanglement between the sender and receiver. We also give an algorithm to determine the circuit for non-binary entanglement-assisted stabilizer quantum codes and some illustrated examples. The codes we constructed do not require the dual-containing constraint, and many non-binary classical codes, like non-binary LDPC codes, which do not satisfy the condition, can be used to c...

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…

GERMAN EPISTOLARY HERITAGE OF IVAN FRANKO AND ITS DOMINANT FEATURES

Directory of Open Access Journals (Sweden)

V.V. Tkachivsky

2015-09-01

Full Text Available The epistolary heritage of Ivan Franko contains over six thousand letters. German epistolary correspondence addressed to Vienna, Berlin, Budapest, Heidelberg, Freiburg, Stanislav, Chernivtsi. The author of the article analyzes also the response letters sent to I. Franko. His first letters in German were addressed to Olha Roshkevych. They were written under the strong impression that the epistolary novel “The Sorrows of Young Werther” by J.W. von Goethe had had on I. Franko. The Ukrainian writer’s correspondence with H. Kanner and I. Singer the editors of the weekly newspaper “Zeit” is an evidence of their close cooperation. I. Franko had close professional ties with V. Adler one of the founders and leaders of the Austrian social-democratic party, the editor of Vienna newspaper “Arbeiterzeitung”, as well as with V. Adler’s wife Emma. I. Franko’s correspondence with a prominent Slavic scholar V. Yagic didn’t contain only private matters discussions. The third part of it is totally scientific which actually makes it valuable for researchers. I. Franko’s epistolary works reflect his artistic writer’s style neither in form nor in content. His letters are characterized by the depth of his thought, consistency of his statements, clearness of the sentence constructions, tendency to critical assessments of phenomena and scientific findings.

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.)

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.)

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’.

The German-Jewish soldier: from participant to victim.

Science.gov (United States)

Penslar, Derek

2011-01-01

The story of German-Jewish soldiers and veterans of World War I illustrates how, under circumstances of inclusion (even if incomplete) rather than vicious persecution, Jewish suffering in wartime, and with it the forms of collective memory and strategies for commemoration of the dead, could closely parallel, even intersect with, the suffering of Germans as a whole. To be sure, the points of intersection were accompanied by points of deflection. Even when Jews served, fought, suffered and died as German soldiers, their interpretations of the war experience, and their communities’ postwar memory and commemorative practices, differed from those of other Germans. In many ways, however, German-Jewish veterans suffered the aftermath of the war as did other Germans; they shared the prevailing fury over war guilt and reparations, and they retained a strong pride in their military service, a pride through which they interpreted the events of 1933–1945.

Insurance dictionary. German-English-French-Greek. Versicherungswoerterbuch. Deutsch-Englisch-Franzoesisch-Griechisch

Energy Technology Data Exchange (ETDEWEB)

Mueller-Lutz, H L [ed.

1984-01-01

Special technical terms used in the world of insurance can hardly be found in general dictionaries. This is a gap which the 'Insurance dictionary' now presented is designed to fill. In view of its supplementary function, the number of terms covered is limited to 1200. To make this dictionary especially convenient for ready reference, only the most commonly used translations are given for each key word in any of the four languages. This dictionary is subdivided into four parts, each containing the translation of the selected terms in the three other languages. To further facilitate the use of the booklet, paper of different colours was used for the printing of the German, English, French and Greek sections. The present volume was developed from a Swedish insurance dictionary (Fickordbok Foersaekring), published in 1967, which - with Swedish as the key language- offers English, French and German translations of the basic insurance terms.

Learning from the Germans? History and Memory in German and European Discourses of Integration

Directory of Open Access Journals (Sweden)

Andrew Beattie

2007-08-01

Full Text Available History and memory appear to be increasingly important to discussions of European values and identity, as exemplified by references to ‘bitter experiences’ and ‘divided pasts’ in the draft EU constitution. The article takes recent suggestions that Europe could learn from German experiences of confronting multiple difficult pasts as its starting point, and considers critically what lessons those German experiences might in fact hold for ‘Europe’. It explores similarities and differences in the two integration contexts and their dominant approaches to, and assumptions about history and public memory. Specifically, it considers debates about the east-west division of the Cold War and about the place of communism and nazism in public memory. Contrary to common assumptions, the article argues that German experiences are not necessarily worth of European emulation.

Arabic Natural Language Processing System Code Library

Science.gov (United States)

2014-06-01

Adelphi, MD 20783-1197 This technical note provides a brief description of a Java library for Arabic natural language processing ( NLP ) containing code...for training and applying the Arabic NLP system described in the paper "A Cross-Task Flexible Transition Model for Arabic Tokenization, Affix...and also English) natural language processing ( NLP ), containing code for training and applying the Arabic NLP system described in Stephen Tratz’s

GERMANISMS IN THE NORTH ČAKAVIAN DIALECT

Directory of Open Access Journals (Sweden)

Marija Turk

2005-01-01

Full Text Available This paper deals with Germanisms in some local idioms of the North Čakavian dialect. Phonological, morphological and word-formation variants of Germanisms are identified. Special attention is paid to the semantic fields and to the spatial, temporal, functional and stylistic stratification of the Germanisms, and to the degree to which they are recognized and used by speakers who belong to different generations.

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...

Contrasts between the environmental restoration challenges posed by uranium mining and milling in the United States and the former German Democratic Republic

International Nuclear Information System (INIS)

Nelson, R.A.; Chernoff, A.R.; Mager, D.; Goldammer, W.

1993-01-01

The former Soviet Union demands for uranium feed materials were primarily met by the East German Republic. A small area 200 km long and 50 km wide in the provinces of Saxony and Thuringia provided more than half of the uranium concentrate processed by the Soviet Union, and used for nuclear weapons development and power generation. With the majority of the ore processed in Germany of an average lower grade than a number of deposits found around the world, the mining and milling resulted in an enormous scale of surface disturbance and quantities of mill tailings concentrated in a relatively small densely populated geographical area. As a result of the re-unification of the two Germanies, all uranium extraction and processing activities were suddenly brought to a halt for economic reasons. The former soviet-East German corporation responsible for the uranium concentrate production was changed into a German state-operated company tasked with the facility decommissioning and environmental restoration. Code-named WISMUT (the German word for Bismuth) during the cold war, this organization was literally changed overnight from a self-sufficient, autonomously operating and state controlled effort into a public works, environmentally conscious corporation

Comparison of energy conservation building codes of Iran, Turkey, Germany, China, ISO 9164 and EN 832

International Nuclear Information System (INIS)

Fayaz, Rima; Kari, Behrouz M.

2009-01-01

To improve the energy efficiency of buildings via compliance to regulation in Iran, Code No. 19 was devised in 1991. The code lacks high level aims and objectives, addressing the characteristics of Iranian buildings. As a consequence, the code has been revised and is not completely implemented in practice, and still remains inefficient. As with any energy coding system, this code has to identify the right balance between the different energy variables for the Iranian climate and way of life. In order to assist improvements to high level objectives of Code 19, this code is compared with ISO 9164, EN 832, German regulation, TS 825 of Turkey and China's GB 50189 to understand how these have adapted international standards to national features. In order to test the appropriateness of Code 19, five case study buildings in Iran are assessed against Code 19 as well as Turkish standard TS 825 and the results are compared. The results demonstrate that Code 19 is efficient in calculations of building envelope, but it needs improvements in the areas of ventilation, gains from internal and solar sources. The paper concludes by offering suggestions for improving the code.

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)

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

Recent trends in coding theory and its applications

CERN Document Server

Li, Wen-Ching Winnie

2007-01-01

Coding theory draws on a remarkable selection of mathematical topics, both pure and applied. The various contributions in this volume introduce coding theory and its most recent developments and applications, emphasizing both mathematical and engineering perspectives on the subject. This volume covers four important areas in coding theory: algebraic geometry codes, graph-based codes, space-time codes, and quantum codes. Both students and seasoned researchers will benefit from the extensive and self-contained discussions of the development and recent progress in these areas.

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....

Study of the accidental risk of the German fast breeder prototype reactor SNR-300

International Nuclear Information System (INIS)

Koeberlein, K.

1983-01-01

A fact-finding committee of the German Federal Parliament in July 1980 recommended to perform a 'risk-oriented study' of the SNR-300, the German 300 MWe fast breeder prototype reactor being under construction at Kalkar. The main aim of this study was to allow a comparative safety evaluation between the SNR-300 and a modern PWR, thus to prepare a basis for a political decision on the SNR-300. Methods and main results of the study are presented in this paper. In the first step of the study six groups of accidents have been identified which may initiate core destruction. By reliability analyses, expected frequency of each group has been estimated. Conditional probabilities for conceivable reactor tank failure modes have been analysed. Tank failure after core destruction leads to release of energy and radioactive material into the containment system. Such accident sequences have been pursued further. Based on a number of core destruction initiators and tank failure modes and various combinations of success and failure states of the containment systems, detailed calculations of different containment scenarious were carried out. From the results of the plant systems analysis, five release categories have been defined. Possible effects of external events and releases of radioactivity from the spent fuel storage pool have also been analysed. In order to quantify the degree of uncertainty of the calculated frequencies, subjective probability distributions of fixed, but inaccurately known quantities have been propagated through the calculations. Using the release categories as input, accident consequences were calculated for the site Kalkar. Though the uncertainty bandwidths for the accident frequencies estimated in the SNR-300 analysis are much wider than for the PWR, the analysis indicates that frequencies of severe accidents, and consequences, are smaller for the SNR-300 than for the PWR as analysed in the 'German Risk Study'. (orig.)

The German-German history of the nuclear power plant Greifswald. Nuclear power between east and west. 2. ed.

International Nuclear Information System (INIS)

Hoegselius, Per

2015-01-01

The historical study covers the chapters The nuclear power plant Greifswald; Lubmin shortly before the ''Wende'' 1989; the German ''Wende''; from the last vote for the ''Volkskammer (parliament of the German Democratic Republic) to the German reunification; Lubmin in reunified Germany; conclusions and perspectives. In the attachment technical data about the reactors WWER-440/W-230 are summarized, including a list of WWERs in the former eastern bloc countries.

German energy policy in deregulated Europe

International Nuclear Information System (INIS)

Kuhnt, D.

2000-01-01

The author argues in favor of a more fact-oriented German energy policy: Firstly, German energy policy must accept the new European framework of a market economy. This means that German utilities must no longer be burdened with the implementation of political objectives. The German power industry needs a level playing field for competition on a European scale. Consequently, also the European partner countries should not limit themselves to the minimum conditions of the Single Market Directive in opening their markets. Secondly, German energy policy must develop new forms of cooperation with the power industry so as to maintain domestic employment and the addition of value despite considerably stronger competitive pressure. Also the conflicting targets of sustainability, continuity of supply, and economic viability must not only be discussed, but must be turned into productive approaches. Thirdly, this means that there must be no inadmissible solution in matters nuclear. If the German power industry is to remain strong, in the interest of domestic jobs and opportunities for the future, it must not lose any more domestic market share to other European companies. Fourthly, we need a new energy policy which takes cognizance of the results of market development in a more rational, less emotional way. In this respect, it should be limited henceforth to supporting renewable energies and technologies so as to enhance energy efficiency in line with market requirements. Fifthly, German energy policy must not commit the mistake of enforcing deregulation and, at the same time, exempting large segments of the market from competition. Thus, the planned expansion of renewable energies, and the increase in cogeneration to more than thirty percent of the German electricty generation, by way of quotas and revenues for electricity from these sources fed into the public grid, are incompatible with competition in Europe. The electricity tax within the framework of the eco tax, the

Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

International Nuclear Information System (INIS)

Keco, M.; Debrecin, N.; Grgic, D.

2005-01-01

Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

RadLex - German version: a radiological lexicon for indexing image and report information

International Nuclear Information System (INIS)

Marwede, D.; Lobsien, D.; Kahn, T.; Daumke, P.; Marko, K.; Schulz, S.

2009-01-01

Purpose: Since 2003 the Radiological Society of North America (RSNA) has been developing a lexicon of standardized radiological terms (RadLex) intended to support the structured reporting of imaging observations and the indexing of teaching cases. The aim of this study was to translate the first version of the lexicon (1 - 2007) into German and to implement a language-independent online term browser. Materials and Methods: RadLex version 1 - 2007 contains 6303 terms in nine main categories. Two radiologists independently translated the lexicon using medical dictionaries. Terms translated differently were revised and translated by consensus. For the development of an online term browser, a text processing algorithm called morphosemantic indexing was used which splits up words into small semantic units and compares those units to language-specific subword thesauri. Results: In total 6240 of 6303 terms (99 %) were translated. Of those terms 3965 were German, 1893 were Latin, 359 were multilingual, and 23 were English terms that are also used in German and were therefore maintained. The online term browser supports a language-independent term search in RadLex (German/English) and other common medical terminology (e.g., ICD 10). The term browser displays term hierarchies and translations in different frames and the complexity of the result lists can be adapted by the user. Conclusion: RadLex version 1 - 2007 developed by the RSNA is now available in German and can be accessed online through a term browser with an efficient search function. This is an important precondition for the future comparison of national and international indexed radiological examination results and the interoperability between digital teaching resources. (orig.)

User's manual for computer code RIBD-II, a fission product inventory code

International Nuclear Information System (INIS)

Marr, D.R.

1975-01-01

The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)