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Sample records for atw schnellstatistik kernkraftwerke

  1. ATW neutronics design studies.

    Energy Technology Data Exchange (ETDEWEB)

    Wade, D. C.; Yang, W. S.; Khalil, H.

    2000-11-10

    The Accelerator Transmutation of Waste (ATW) concept has been proposed as a transuranics (TRU) (and long-lived fission product) incinerator for processing the 87,000 metric tonnes of Light Water Reactor used fuel which will have been generated by the time the currently deployed fleet of commercial reactors in the US reach the end of their licensed lifetime. The ATW is proposed to separate the uranium from the transuranics and fission products in the LWR used fuel, to fission the transuranics, to send the LWR and ATW generated fission products to the geologic repository and to send the uranium to either a low level waste disposal site or to save it for future use. The heat liberated in fissioning the transuranics would be converted to electricity and sold to partially offset the cost of ATW construction and operations. Options for incineration of long-lived fission products are under evaluation. A six-year science-based program of ATW trade and system studies was initiated in the US FY 2000 to achieve two main purposes: (1) ''to evaluate ATW within the framework of nonproliferation, waste management, and economic considerations,'' and (2) ''to evaluate the efficacy of the numerous technical options for ATW system configuration.'' This paper summarizes the results from neutronics and thermal/hydraulics trade studies which were completed at Argonne National Laboratory during the first year of the program. Core designs were developed for Pb-Bi cooled and Na cooled 840 MW{sub th} fast spectrum transmuter designs employing recycle. Additionally, neutronics analyses were performed at Argonne for a He cooled 600 MW{sub th} hybrid thermal and fast core design proposed by General Atomics Co. which runs critical for 3/4 and subcritical for 1/4 of its four year once-thin burn cycle. The mass flows and the ultimate loss of transuranic isotopes to the waste stream per unit of heat generated during transmutation have been calculated on a

  2. Kernkraftwerke

    Science.gov (United States)

    Allelein, H. J.

    Die Nutzung der Kernenergie ist untrennbar mit dem Element Uran verknüpft. Da die Umwandlung eines Elements in ein anderes auf chemischem Weg nicht möglich ist, muss das heute vorhandene Uran durch kosmische Prozesse vor Entstehung der Erde entstanden sein. Reines Uran ist ein silberweiß glänzendes, relativ weiches Schwermetall. In der Elementhäufigkeit steht Uran vor Gold, Silber oder Quecksilber.

  3. Trace Assessment for BWR ATWS Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

    2010-04-22

    A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtained from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.

  4. BWR MOX core monitoring at Kernkraftwerk Gundremmingen

    Energy Technology Data Exchange (ETDEWEB)

    Noel, Alejandro [Studsvik Scandpower (Suisse) GmbH, Nussbaumen AG (Switzerland); Holzer, Robert [NIS Ingenieurgesellschaft GmbH, Alzenau (Germany); Anton, Gerd [Studsvik Scandpower GmbH, Norderstedt (Germany); Smith, Kord [Studsvik Scandpower Inc., Idaho Falls (United States)

    2008-07-01

    The replacement of the core monitoring system for twin KWU Boiling Water Reactors (BWR) is presented. The reactors, Kernkraftwerk Gundremmingen B and C (KGG), are located in Germany. Core monitoring for KGG is more challenging than for most BWR reactors due to its core composition with about 30% MOX fuel assemblies. The objectives of this paper are to discuss the specific MOX modelling aspects in CASMO-4/Simulate-3, the impact of the MOX fuel on several core monitoring aspects like the LPRM detector modelling and to present some core monitoring results since the beginning of GARDEL's operation. The available core monitoring results confirm the accuracy of the underlying physical methods. The core monitoring system replacement att KGG was a common project of Studsvik Scandpower and NIS Ingenieurgesellschaft GmbH, where Studsvik Scandpower supplied its standard core monitoring system GARDEL and NIS was responsible for the computer hardware, system integration and plant specific add-ons. (authors)

  5. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  6. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group.

    Energy Technology Data Exchange (ETDEWEB)

    Collins, E.; Duguid, J.; Henry, R.; Karell, E.; Laidler, J.; McDeavitt, S.; Thompson, M.; Toth, M.; Williamson, M.; Willit, J.

    1999-08-12

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD&D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years.

  7. Study on Fabrication of Ni-5 at.%W Tapes for Coated Conductors from Cylinder Ingots

    DEFF Research Database (Denmark)

    Ma, L.; Suo, H. L.; Yue, Zhao

    2015-01-01

    Ni-5 at.%W (Ni5W) tapes with a strong cube texture were fabricated using the RABiTS technique and by starting from cylindrical shaped ingots. In contrast to a conventional cuboid-shaped ingot, a cylinder shaped ingot has no anisotropy along the axial direction and the resulting tape will therefore...

  8. German nuclear power plants: 1994 performance. Excerpt from the report of the ABE Committee; Deutsche Kernkraftwerke: Betriebsergebnisse 1994. Auszug aus dem Bericht des ABE-Ausschusses

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1995-06-01

    The compact survey of the performance of the nineteen nuclear power plants in operation in Germany covers the year 1994, starting with a general survey and information and continuing with detailed, separate reports on the various generating units. The data include the 1993 operating results. (UA) [Deutsch] Im vorliegenden Bericht werden die Betriebsergebnisse des Jahres 1994 der in der Bundesrepublik Deutschland betriebenen Kernkraftwerke zusammenfassend dargestellt. Nach einem einleitenden Ueberblick werden abschnittsweise die einzelnen Kernkraftwerke behandelt, wobei die Betriebsergebnisse 1993 wiedergegeben werden. (UA)

  9. Accelerator technology for the Los Alamos ATW (accelerator transmutation of nuclear waste) system

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, G.P.

    1991-01-01

    The Los Alamos concept for accelerator transmutation of nuclear waste (ATW) employs a high-power proton linear accelerator to generate intense fluxes of thermal neutrons (>10{sup 16} n/cm{sup 2}-s) through spallation on a lead-bismuth target. The nominal beam energy for an ATW accelerator is 1.6 GeV, with average current requirements ranging from 250 mA to 30 mA, depending on application specifics. A recent study of accelerator production of tritium (APT) led to the development of a detailed point design for a 1.6 GeV, 250 mA cw proton linac. The accelerator design was reviewed by the Energy Research Advisory Board (ERAB) and found to be technically sound. The Panel concluded that linac of this power level could now be implemented within the existing technology base, given an adequate component development program and an integrated engineering demonstration of the front end.

  10. Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available.

  11. The feasibility study I on the blanket fuel options for the ATW/HYPER

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok; Meyer, M.K; Hayes, S.L

    2001-01-01

    The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended.

  12. LOFT L9-3 ATWS Experiment Simulation using the SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chang Keun; Lee, Dong Hyuk; Kim, Yo Han; Ha, Sang Jun [KEPCO Research Institute, Daejeon (Korea, Republic of); Hong, Sung Yull [bYeungnam University, Gyeongsan (Korea, Republic of)

    2011-05-15

    The Korea nuclear industry has developed a best estimated two-phase three-filed thermal-hydraulic analysis code, SPACE(Safety and Performance Analysis Code for Nuclear Power Plants), for safety analysis and design of a PWR(Pressurized Water Reactor). As the first phase, the demo version of SPACE code was released on March, 2010. And the code has been verified and improved according to the Validation and Verification (V and V) matrix prepared for the SPACE code as the second phase of the development. In this study, the LOFT (Loss Of Fluid Test) L9-3 Anticipated Transient Without Scram (ATWS) experiment has been simulated using the SPACE code as one of the V and V work. The results were compared with those of the experiment and other code simulation

  13. CSAU methodology and results for an ATWS event in a BWR using information theory methods

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J.L., E-mail: jlcobos@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Escrivá, A., E-mail: aescriva@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Mendizabal, R., E-mail: rmsanz@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Pelayo, F., E-mail: fpl@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Melara, J., E-mail: jls@iberdrola.es [IBERINCO, IBERDROLA Ingeniería y Construcción, Madrid (Spain)

    2014-10-15

    Highlights: • We apply the CSAU methodology to an ATWS in a BWR using information theory methods. • We show how to perform the selection of the most influential inputs on the critical safety parameter. • We apply the maximum entropy principle to get the input parameter distribution. • We examine the maximum relative entropy principle to update the input parameter PDF. • We quantify the uncertainty of the critical safety parameter using order statistics and information theory. - Abstract: This paper shows an application of the CSAU methodology to an ATWS in a BWR reactor, when the temperature of the suppression pool is taken as the critical safety parameter. The method combines CSAU methodology with recent techniques of information theory. In this paper we use auxiliary tools to help in the evaluation and improvement of the parameters distribution that enter in the elements II and III of CSAU based methodologies. These tools have been implemented in two FORTRAN programs: GEDIPA (Generation of the Parameter Distribution) and UNTHERCO (Uncertainty in Thermal Hydraulic Codes). The first one analyzes the information data available on a given parameter or parameters with the goal to know all the information about the probability distribution function of these parameters. The second apply information theory methods, as the maximum entropy principle (MEP) and the maximum relative entropy Principle (MREP), in order to build conservative distribution functions for the parameters from the available data. Also, the distribution function of a given parameter can be updated using the MREP principle when new information is provided. UNTHERCO performs the MONTECARLO sampling for a given set of parameters when the distribution function of these parameters is previously known. If the distribution of a parameter is unknown, then, the MEP is applied to deduce the distribution function for this parameter.

  14. Decommissioning of the Wuergassen nuclear power plant, a commercial challenge; Stillegung Kernkraftwerk Wuergassen, eine kommerzielle Aufgabe

    Energy Technology Data Exchange (ETDEWEB)

    Albrecht, S. [Kernkraftwerk Wuergassen (Germany); Schiffer, K.J. [Kernkraftwerk Wuergassen (Germany)

    1996-06-01

    In response to the inspection results which detected cracks in the core shroud, economic aspects have induced PreussenElektra to opt for decommissioning and dismantling of the Wuergassen reactor. As this shutdown of the nuclear power plant is not a planned shutdown, costs arising in addition to the original decommissioning framework studies have to be assessed, especially the expenditure for the adjusted plant manpower requirements, and the additional operating and phase-out costs. Experience has shown that the decommissioning of a nuclear power plants does not pose problems in terms of safety or technology, but still is a commercial challenge. Expense forecasts have to be adjusted in response to the unplanned shutdown. PreussenElektra therefore has set up a modified project and operating structure. The analysis and evaluation of the first decommissioning phase will show whether the cost assessment approaches are in agreement with reality. (orig.) [Deutsch] PreussenElektra hat aufgrund der Risse im Kernmantel und der damit verbundenen wirtschaftlichen Betrachtungen die Stillegung und den Direkten Rueckbau des Kernkraftwerkes Wuergassen beschlossen. Da es sich bei der Stillegung des KWW um eine ungeplante Stillegung handelt, sind ueber die urspruenglichen Studien hinaus zusaetzliche Kostenbloecke zu betrachten. Insbesondere sind hierbei die Kosten fuer den angepassten Eigenpersonaleinsatz und die zusaetzlichen Betriebs- und Auslaufkosten zu erwaehnen. Wie die Erfahrungen zeigen ist die Stillegung von kerntechnischen Anlagen technisch und sicherheitstechnisch kein Problem. Es bleibt die Herausforderung der Wirtschaftlichkeit. Die bisherigen Kostenprognosen muessen aufgrund der ungeplanten Stillegung angepasst werden. Der Herausforderung tritt PreussenElektra mit einer angepassten Projekt- und Betriebsstruktur entgegen. Die Auswertung der ersten Stillegungsphase wird zeigen, ob die vorgestellten Ansaetze die Realitaet richtig beschreiben koennen. (orig.)

  15. Evolution of microstructure, texture and topography during additional annealing of cube-textured Ni–5at.%W substrate for coated conductors

    DEFF Research Database (Denmark)

    Wulff, Anders Christian; Mishin, Oleg; Grivel, Jean-Claude

    2012-01-01

    Microstructure, texture and topography have been studied in a recrystallized Ni–5at.%W substrate before and after additional annealing at 1025C for 1 h. The initial recrystallized material contained a strong cube texture and a high fraction of low angle grain boundaries. R3 boundaries were also f...

  16. Highly textured Gd2Zr2O7 films grown on textured Ni-5 at.%W substrates by solution deposition route: Growth, texture evolution, and microstructure dependency

    DEFF Research Database (Denmark)

    Yue, Zhao; Grivel, Jean-Claude; Napari, M.

    2012-01-01

    or crystallization in the thicker films. This work not only demonstrates a route for producing textured Gd2Zr2O7 buffer layers with dense structure directly on technical substrates, but also provides some fundamental understandings related to chemical solution derived films grown on metallic substrates.......Growth, texture evolution and microstructure dependency of solution derived Gd2Zr2O7 films deposited on textured Ni-5 at.%W substrates have been extensively studied. Influence of processing parameters, in particular annealing temperature and dwell time, as well as thickness effect on film texture...... the difference of interfacial energy along two directions in the anisotropic metallic substrate. Growth of Gd2Zr2O7 films displays an ultrafast kinetics under optimized conditions. Independency of sharp epitaxial (004) and polycrystalline (222) orientation is revealed from further synchrotron diffraction studies...

  17. Impacts of the amended radiological protection regulations for clearance and exemption from regulatory control on dismantling of a nuclear power plant; Wechselwirkung zwischen den neuen Freigaberegelungen und dem Rueckbau eines Kernkraftwerks: Erwartungen und Erfahrungen

    Energy Technology Data Exchange (ETDEWEB)

    Pollmann, E. [PreussenElektra Kernkraft GmbH und Co. KG, Kernkraftwerk Wuergassen, Beverungen (Germany)

    2000-07-01

    The fundamental objectives of the Wuergassen power plant dismantling project relate to : - Ensuring efficient general and radiological protection at work throughout project activities,- optimizing economic efficiency,- maintaining a continuous, rapid mass flow in waste management activities. The project for dismantling of the Wuergassen reactor is of significance as kind of a pioneering project, to establish experience for future reactor dismantling activities in Germany. (orig./CB) [German] Die uebergeordneten Ziele des gesamten Projektes sind: - Allgemeiner und radiologischer Arbeitsschutz, - Optimierung der Wirtschaftlichkeit, - hoher, kontinuierlicher Massenfluss zur Entsorgung. Der Rueckbau des Kernkraftwerkes Wuergassen hat Prototyp-Charakter fuer industriellen Rueckbau von Kernkraftwerken in Deutschland. (orig./SR)

  18. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133{degrees}F) has a 95-percentile uncertainty of 14.4 K (26{degrees}F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175{degrees}F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6{degrees}F).

  19. German nuclear power plants: 1995 performance data. Excerpt from the report of the ABE-Committee; Deutsche Kernkraftwerke: Betriebsergebnisse 1995. Auszug aus dem Bericht des ABE-Ausschusses

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1996-05-01

    The 1995 Annual Report published by the Technical Committee for the Exchange of Operating Experience (ABE Committee) within the Technical Association of Operators of Large Power Plants (VGB) contains a summary record of the performance of the nuclear power plants operated in Germany. The nineteen nuclear generating units operated in Germany in 1995 (Muelheim-Kaerlich was down also throughout 1995, as was Wuergassen), with a cumulated installed capacity of 22,063 MWe, converted a total of 154,144 GWh of nuclear power into electricity, which is approx. 2% more than the 151,163 GWh of 1994. The share contributed to the public electricity supply amounted to 33%. The ABE report also mentions the AVR experimental nuclear power station in Juelich, which is currently being decommissioned. (orig.) [Deutsch] Im Jahresbericht 1995 des Fachausschusses fuer Austausch von Betriebserfahrungen (ABE-Ausschuss) in der VGB Technischen Vereinigung der Grosskraftwerksbetreiber werden die Betriebsergebnisse der in Deutschland betriebenen Kernkraftwerke dargestellt. Die 19 im Jahre 1995 in Deutschland in Betrieb gewesenen Kernkraftwerksbloecke (Muelheim-Kaerlich war auch 1995 nicht in Betrieb, Wuergassen war das ganze Jahr ueber abgeschaltet und wird endgueltig stillgelegt) mit einer installierten Leistung von 22 063 MWe (brutto) haben insgesamt 154 114 GWh aus Kernenergie in elektrische Arbeit umgewandelt. 2% mehr als 1994 mit 151 163 GWh. Der Anteil an der oeffentlichen Versorgung betrug wie in den Vorjahren seit 1988 etwa ein Drittel, der Anteil am Primaerenergiebedarf rd. 10%. Der ABE-Bericht geht ferner ein auf das in der Stillegung befindliche Versuchskernkraftwerk der Arbeitsgemeinschaft Versuchsreaktor (AVR) in Juelich. (orig.)

  20. 1996: German nuclear power plants` performance data. Excerpt from the report of the ABE Committee; 1996: Betriebsergebnisse deutscher Kernkraftwerke. Auszug aus dem Bericht des ABE-Ausschusses

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1997-05-01

    The 1996 Annual Report published by the Technical Committee for the Exchange of Operating Experience (ABE Committee) within the Technical Association of Operators of Large Power Plants (VGB) contains a summary record of the performance of the nuclear power plants operated in Germany. The nineteen nuclear generating units operated in Germany in 1996 (Muelheim-Kaerlich was down also throughout 1996, as was Wuergassen), with a cumulated installed capacity of 22,149 MWe, converted a total of 161,702 GWh of nuclear power into electricity, which is approx. 4,9% more than the 154,144 GWh of 1995. The share contributed to the public electricity supply amounted to 33%. The ABE report also mentions the AVR experimental nuclear power station in Juelich, which is currently being decommissioned. (orig.) [Deutsch] Im Jahresbericht 1996 des Fachausschusses fuer Austausch von Betriebserfahrungen (ABE-Ausschuss) in der VGB Technischen Vereinigung der Grosskraftwerksbetreiber werden die Betriebsergebnisse der in Deutschland betriebenen Kernkraftwerke dargestellt. Die 19 im Jahre 1996 in Deutschland in Betrieb gewesenen Kernkraftwerksbloecke (Muelheim-Kaerlich war auch 1996 nicht in Betrieb, Wuergassen wird endgueltig stillgelegt) mit einer installierten Leistung von 22 149 MWe (brutto) haben insgesamt 161 702 GWh aus Kernenergie in elektrische Arbeit umgewandelt, 4,9% mehr als 1995 mit 154 144 GWh. Der Anteil an der oeffentlichen Versorgung betrug wie in den Vorjahren seit 1988 etwa ein Drittel, der Anteil am Primaerenergiebedarf rd. 10%. Der ABE-Bericht geht ferner ein auf das in der Stillegung befindliche Versuchskernkraftwerk der Arbeitsgemeinschaft Versuchsreaktor (AVR) in Juelich. (orig.)

  1. Official notice concerning a licence issued in compliance with paragraph 15, sub-sec. 3 and paragraph 17 of the Nuclear Installations Ordinance, for modification of the Wuergassen nuclear power plant, issued on 23 September 1994, representing the 4th supplementing permit to licensing notice No. 7/10 KWW. As of 11 November 1994; Hinweis auf die oeffentliche Bekanntmachung gemaess Para. 15 Abs. 3 und Para. 17 der Atomrechtlichen Verfahrensverordnung (AtVfV) ueber eine Aenderungsgenehmigung vom 23. September 1994 fuer das Kernkraftwerk Wuergassen 4. Ergaenzung zum Bescheid Nr. 7/10 KWW. Vom 11. November 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-11-19

    Licence for the erection of a new filtering system building and a filtering system for optional use, to accept radwaste and radioactive remnants for interim storage, and for the installation of two cranes in the nuclear power plant. [Deutsch] Genehmigung zum Bau eines neuen Filtergebaeudes und einer Bedarfsfilteranlage, zur Lagerung radioaktiver Abfaelle und Reststoffe, sowie zur Errichtung von 2 Krananlagen im Kernkraftwerk

  2. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Fourth quarterly report 1997; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. 4. Quartal 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-08-01

    The report presents a brief survey of notifiable events in German nuclear power plants and research reactors of the given output category, covering the last quarter of the year 1997. (orig./CB) [Deutsch] Der vorliegende Bericht enthaelt die Uebersicht ueber die meldepflichtigen Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen (Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet) der Bundesrepublik Deutschland fuer das vierte Vierteljahr 1997. (orig./AJ)

  3. Development of One Meter Long Double-Sided CeO2 Buffered Ni-5at.%W Templates by Reel-to-Reel Chemical Solution Deposition Route

    DEFF Research Database (Denmark)

    Yue, Zhao; Konstantopoulou, K.; Wulff, Anders Christian

    2013-01-01

    . The major achievement of the design is to combine the dip coating and drying processes in order to overcome the technical difficulties of dealing with the wet films on both sides of the tape. We report the successful application of the design to fabricate a one-meterlong double side coated CeO2/Ni − 5at......%W template. The CeO2 films on both sides exhibit a dense, crack-free morphology, and a high fraction of cube texture on the surface. Homogeneity studies on global texture over the length also reveal that the average full width at half maximum values of the in-plane and out-of-plane orientation on the CeO2...... layer are 7.2◦ and 5.8◦ with standard deviation of 0.26◦ and 0.34◦, respectively, being indicative of the high quality epitaxial growth of the films prepared in the continuous manner. An all chemical solution derived YBCOLow−TFA/Ce0.9La0.1O2/Gd2Zr2O7/CeO2 structure is obtained on a short sample...

  4. Nuclear power. BGH ruling of 16.1.1997 (Az: III ZR 117/95) relating to the action for damages in the matter of the shut-down Muehlheim-Kaerl-Kaerlich Reactor; Kernenergie. BGH-Urteil vom 16.1.1997 (Az.: III ZR 117/95) zum Schadensersatzprozess wegen des Kernkraftwerks Muelheim-Kaerlich

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1997-04-01

    The German Federal High Court (BGH) non-appealably rejected part of the claims for damages of the owner/operator of the Muehlheim-Kaerlich nuclear power station asserted in an action against the Federal State of Rhineland-Palatinate and remanded the case to the lower court for clarification of the remaining claims. The plant operator claimed compensation for damage incurred in the wake of the annulment of the first partial permit issued for the nuclear power plant in 1975 and declared to be void for reasons of non-compliance with the licensing provisions of section 7, sub-section 2 AtG (Atomic Energy Act), and breach of official duty of civil servants of the licensing authority of Rhineland-Palatinate. Due to this decision of the Federal Administrative Court, the plant was shut down in 1988. The claims asserted by the operator relate among other items to compensation for construction, operation and shut-down operation costs. (orig./CB) [Deutsch] Der Bundesgerichtshof hat einen Teil der Schadensersatzansprueche der Betreiberin des Kernkraftwerks Muehlheim-Kaerlich gegen das Land Rheinland-Pfalz endgueltig abgewiesen und zur Klaerung der verbleibenden Ansprueche den Rechtsstreit an die Vorinstanz zurueckverwiesen. In diesem Verfahren ging es um Schadensersatz infolge der Aufhebung der ersten Teilgenehmigung fuer das Kernkraftwerk aus dem Jahre 1975, wegen Nichterfuellung der Genehmigungsvoraussetzungen des paragraph 7 Abs. 2 AtG und Amtspflichtverletzung der Genehmigungsbehoerde des Landes. Aufgrund der nachfolgenden Stillegung des Kraftwerks 1988 verlangte die Betreiberin Schadensersatz u.a. fuer ihren Errichtungs- und Betreibungsaufwand und die Betriebskosten fuer den Stillegungsbetrieb. (orig./CB)

  5. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Second quarterly report 1998; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Vierteljahresbericht 2. Quartal 1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-31

    The report contains the documentation of notifiable events in the defined reactors recorded over the second quarter of 1998. The documentation is prepared according to the national notification and reporting system prescribed by the relevant law in Germany, and is filed to the national atomic energy supervisory authorities in Germany for documentation in the national record. (orig./CB) [Deutsch] Der vorliegende Bericht enthaelt die Uebersicht ueber die meldepflichtigen Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen (Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistuung 50 kW thermische Dauerleistung ueberschreitet) der Bundesrepublik Deutschland fuer das zweite Vierteljahr 1998. Meldepflichtige Ereignisse in Kernkraftwerken der Bundesrepublik Deutschland werden seit 1975 nach bundeseinheitlichen Meldekriterien in der jeweils gueltigen Fassung an die atomrechtlichen Aufsichtsbehoerden gemeldet und in einer zentral gefuehrten Liste erfasst. (orig.)

  6. Expert report of ENSI on the request of KKN AG for a general license - Project 'New nuclear power plant Niederamt'; Gutachten des ENSI zum Rahmenbewilligungsgesuch der KKN AG. Neubauprojekt Kernkraftwerk Niederamt

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-09-15

    The 'Kernkraftwerk Niederamt AG' (KKN) Company submitted to the Swiss Federal Inspectorate of Nuclear Safety (ENSI) a request for a general license for a new power plant to be built near to the Goesgen power plant. According to the law, all damage risks with a probability higher than 10{sup -4}/a must be taken into account through protection measures. The considered risks concern the power plant itself as well as the population in the neighbourhood and the environment. The purpose of the general license is to demonstrate that the site chosen for the foreseen power plant is acceptable and that the risks can be counteracted through adequate measures. The buildings of the power plant and their partition over the two banks of the Aare River are briefly described. The reactor is a Light Water Reactor of third generation with a maximum thermal power of 5.8 GW{sub th}. The main cooling is provided by a hybrid system of water evaporation and air heating, what reduces the plume at the exit of the cooling tower. First, it is demonstrated that, in the case of a very unlikely severe accident in the power plant, the people in the neighbourhood can be evacuated quickly. Then, numerous types of possible accidents in the neighbourhood of the power plant are analyzed in order to settle their possible negative influence on the operation of the power plant: bursting of gas containers on the neighbouring roads and railways, fires of all types of hydrocarbons, air pollution through chloride gas, etc. The check by ENSI of the KKN studies on the potential danger for the power plant through neighbouring industrial plants, roads or railways demonstrated that none of the considered accidents presents an unacceptable risk for the power plant: on the one hand, these plants are located too far from the power plant, so that a sensible injury to the power plant safety can be excluded; on the other, the protection of the power plant can be guaranteed through appropriate technical

  7. The Human Factors System of VGB. The body of measures implemented by the NPP operating companies for HF management and optimisation of the man-machine interface; VGB-Human-Factors-System. Human-Factors-Massnahmen der Kernkraftwerks Betreiber zur Optimierung der Mensch-Maschine-Schnittstelle

    Energy Technology Data Exchange (ETDEWEB)

    Eisgruber, H.; Janssen, G. [RWE Energie AG, Essen (Germany)

    1999-04-01

    One way of ensuring high human reliability is to systematically record, analyse and optimise the identified human factors influencing the safety of operation of NPPs. There are two kinds of human factors to be considered: Those identified and characterised in their actual influence on the man-machine system after an event has happened, and the potential human factors, as far as they are known. The purpose of the paper presented is to: - explain the man-machine interface on the basis of the descriptions delivered by work science, as well as with the information model (acquisition, processing and application of information); - explain their influence on the human performance; - identify the organisational units/competences of a nuclear power plant with respect to responsibility for dealing with those factors influencing human performance; - describe suitable joint action of the responsible organisational units; - present results of HF system reviews and modifications. (orig./CB) [German] Das hohe Mass an menschlicher Zuverlaessigkeit wird u.a. durch eine systematische Erfassung, Analyse und Optimierung menschlicher Einflussfaktoren auf Bedienvorgaenge im Kernkraftwerk sichergestellt. Es sind einerseits die Einflussfaktoren, die tatsaechlichen in einem Mensch-Maschine-System wirksam geworden sind und andererseits die Einflussfaktoren, die potentiell vorhanden sein koennen. Ziel dieser Abhandlung ist es: - Die Mensch-Maschine-Schnittstelle auf Basis ihrer arbeitswissenschaftlichen Definition und des Informationsmodells (Informationsaufnahme, Informationsverarbeitung, Informationsumsetzung) fuer Praktiker im Kraftwerk zu erlaeutern und die Einfluesse auf die menschliche Leistung beispielhaft darzulegen. - Abteilungen/Zustaendigkeiten des Kraftwerkes fuer die Behandlung von Einfluessen auf die menschliche Leistungsfaehigkeit aufzuzeigen. - Das Zusammenwirken der zustaendigen Abteilungen im Kraftwerk bei der Behandlung von leistungsbeeinflussenden Faktoren zu beschreiben

  8. Nuclear data review and compilation for ATW systems

    Energy Technology Data Exchange (ETDEWEB)

    Guzhovskii, B.; Gorelov, V.; Il`in, V.; Farafontov, G.; Grebennikov, A.

    1994-10-01

    In order to solve the problem of nuclear power waste transmutation in neutron flux it is necessary to know the characteristics of neutron interaction for a great number of nuclei in the energy range from 0 to hundreds of MeV. The authors distinguished the most important aspect of this problem that one of nuclear data for actinides, (from Th to Cm isotopes) They have given the overview of evaluations of characteristic of interaction between neutrons and these nuclei leading to transformation from target-nucleus to neighboring actinide-nucleus or fission fragments in the limited energy range from 0 to 14 MeV. The review was carried out by comparison of mentioned characteristics from the modern versions of ENDL-82, JENDL-3, ENDF/B-6 and BROND-2 neutron evaluated data among themselves and with recommended data of previous publications and, in some cases, with the measurement results. ENDL-82 and ENDF/B-6 versions were made in USA laboratories, JENDL-3 was made in the laboratories of Japan and BROND-2 version was made in the laboratories of former USSR. The comparison of nuclear data from various libraries was carried out by the most economic method permitting, nevertheless, fully judge of available uncertainties in the knowledge of competitive nuclear data which are important from the point of view of problem of transmutation in various energies neutron flux. The following characteristics were considered: (a) fission and capture cross-sections at thermal point (E{sub n}=0.0253 eV); (b) infinitely dilute resonance integrals of fission and capture designated by I{sub f} and I{sub {gamma}} (c) averaged on {sup 252}Cf spontaneous fission neutron spectrum cross-sections of fission, capture and the (n,2n) reactions; (d) cross-sections of fission and the (n,2n), (n,3n) reactions at the point En = 14 MeV; (e) fission and capture resonance integrals for a interval of sets with the increasing upper (E {sub max}) and lower (E {sub min}) limits of integral.

  9. Nuclear power plant with a containment. Kernkraftwerk mit einer Sicherheitshuelle

    Energy Technology Data Exchange (ETDEWEB)

    Barthelmes, C.P.

    1982-03-25

    In nuclear power plants there is usually a containment incorporating components bearing activity. If in the cladding free hydrogen develops, controlled oxidation must be ensured by means of a recombination device, in order to prevent oxyhydrogen explosions. For this purpose, a permanent thorough mixing of the gases in the containment is required. This can be achieved by vertical shafts reaching to at least half the height of the containment and provided with heating devices to initiate the gas circulation by the stack effect. These heating devices mainly serve as a thermal recombinator.

  10. Sicherheitsanforderungen für zukünftige Kernkraftwerke

    OpenAIRE

    Kugeler, K.; Phlippen, R. H.; Alkan., Z.; Kugeler, M.

    2000-01-01

    Presently, about 450 different nuclear power plants, partly highly developed, are operating worldwide. Thereby a great contribution to the worldwide current supply is made by nuclear energy. With a capacity amounting to 360 GWI, these nuclear plants meet 17% of the total energy demand. The use of nuclear energy will become more and more important on a longterm basis, if attention is paid to aspects of resource saving, economic efficiency and care of the environment, with special regard to the...

  11. Analysis of loss of off-site power ATWS in VVER-440 concept

    Energy Technology Data Exchange (ETDEWEB)

    Hoeppner, G.; Siltanen, P.; Kotro, J.

    1987-01-01

    During 1985 the Finnish state-owned utility Imatran Voima Oy signed a work order with Gesellschaft fuer Reaktorsicherheit mbH of the Federal Republic of Germany (GRS) for the analysis of abnormal transients in a pressurized water reactor (PWR) concept based on a Soviet design. The results of these calculations were intended to be introduced into the licensing process and to support a decision to build such a nuclear power station. A computer model was constructed of the VVER-440 concept, a 500-MW(electric) PWR designed in the USSR and modified for Finland. The ALMOD4 code, developed at GRS, was used for the investigation. The ALMOD4 code is a fast running code for the analysis of operational and abnormal transients in PWRs. Input data were set up to calculate anticipated transients without scram, most notably the loss of off-site power case. One-dimensional neutron kinetics was used to correctly model the neutronics feedback of axially distributed moderator density and fuel temperature in a changing axial power profile. Interlocking signals and the engineered safety systems were modeled to assess the overall systems response to this abnormal transient. Special analytical problems were encountered since a detailed and verified model of the steam generator (SG) with horizontally positioned heat exchanger tubes was not available. Therefore, two bounding calculations were performed with different SG models.

  12. Fabrication of the Textured Ni-9.3at.%W Alloy Substrate for Coated Conductors

    DEFF Research Database (Denmark)

    Gao, M. M.; Suo, H. L.; Grivel, Jean-Claude

    2011-01-01

    It is difficult to obtain a sharp cube texture in the Ni-9.3at.% W substrate used for coated conductors due to its low stacking fault energy. In this paper, the traditional cold rolling procedure was optimized by introducing an intermediate recovery annealing. The deformation texture has been...

  13. KWL Lingen nuclear plant. Technical annual report 2016; KWL Kernkraftwerk Lingen. Technischer Jahresbericht 2016

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2017-07-01

    The technical annual report 2016 for KWL (Lingen nuclear plant) covers the following sections: dismantling project management and operation, monitoring and clearance; waste management, technical qualification, security and safety, central tasks; licensing and supervision procedures, operational data, radiation monitoring, radioactive materials, in-service inspections.

  14. Safety culture in nuclear power plants. Proceedings; Sicherheitskultur im Kernkraftwerk. Seminarbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-01

    As a consequence of the INSAG-4 report on `safety culture`, published by the IAEA in 1991, the Federal Commission for the Safety of Nuclear Power Plants (KSA) decided to hold a one-day seminar as a first step in this field. The KSA is an advisory body of the Federal Government and the Federal Department of Transport and Energy (EVED). It comments on applications for licenses, observes the operation of nuclear power plants, assists with the preparation of regulations, monitors the progress of research in the field of nuclear safety, and makes proposals for research tasks. The objective of this seminar was to familiarise the participants with the principles of `safety culture`, with the experiences made in Switzerland and abroad with existing concepts, as well as to eliminate existing prejudices. The main points dealt with at this seminar were: - safety culture from the point of view of operators, - safety culture from the point of view of the authorities, - safety culture: collaboration between power plants, the authorities and research organisations, - trends and developments in the field of safety culture. Invitations to attend this seminar were extended to the management boards of companies operating Swiss nuclear power plants, and to representatives of the Swiss authorities responsible for the safety of nuclear power plants. All these organisations were represented by a large number of executive and specialist staff. We would like to express our sincerest thanks to the Head of the Federal Department of Transport and Energy for his kind patronage of this seminar. (author) figs., tabs., refs.

  15. Nuclear power station with a water-cooled reactor pressure vessel. Kernkraftwerk mit einem wassergekuehlten Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, R.; Brunner, G.; Jost, N.

    1987-10-29

    Nuclear radiation produces radiolysis gases, which are undesirable for corrosion and oxyhydrogen gas reasons. To limit the proportion of this radiolysis gas, the invention provides that catalytic surfaces should be introduced into the primary circuit, to produce recombination of hydrogen and oxygen. These surfaces can be accommodated in the upper part of the reactor pressure vessel. The live steam screen can also have a catalytic surface.

  16. German nuclear power plants: Performance in 1993. Deutsche Kernkraftwerke: Betriebsergebnisse 1993

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1994-05-01

    The compact survey of the performance of the nineteen nuclear power plants in operation in Germany covers the year 1993, starting with a general survey and information and continuing with detailed, separate reports on the various generating units. (UA)

  17. German nuclear power plants: 1992 performance data. ABE report. Deutsche Kernkraftwerke: Betriebsergebnisse 1992. ABE-Bericht

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1993-05-01

    The report presents the 1992 performance data of the nuclear power plants in operation in the Federal Republic of Germany. The data include information on incidents of relevance to safety of operation, important modifications, and radioactivity emissions in the reporting year. The individual nuclear power plant data also include the 1992 operating diagrams. (orig.)

  18. MORECA-GT: Interactive simulator for gas-turbine modular HTGR transients and heatup accidents with ATWS options

    Energy Technology Data Exchange (ETDEWEB)

    Ball, S.J.; Nypaver, D.J.

    1994-03-01

    An interactive simulation code for studying postulated heatup accidents in modular high-temperature gas-cooled reactors (MHTGRs) has been adapted to assist with parametric design studies of the US Department of Energy`s (DOE`s) direct-cycle gas-turbine MHTGR concept. The studies show that the proposed MHTGR designs are very robust and can generally withstand the consequences of extremely low probability accidents with little or no damage to the reactor`s fuel or metallic components.

  19. 78 FR 5864 - Waiver of Aeronautical Land-Use Assurance: Outagamie County Regional Airport (ATW), Appleton, WI

    Science.gov (United States)

    2013-01-28

    ... Training Center (PSTC) by the Fox Valley Technical College (FVTC). The PSTC is an educational campus... Layout Plan (ALP) dated January 13, 1993, and the Exhibit ``A'' property map. This parcel, as shown on the ALP, is not needed for aeronautical use. There are no impacts to the airport by allowing it to...

  20. Proceedings of the 6th Annual Advance Technology Workshop. ATW’ 98. 19-20th of May, 1998

    Science.gov (United States)

    1998-05-01

    References [Dumeur 94] : R.Dumeur "Synthese de Comportements des Animaux Individueis et Collectifs par Algorithmes genetiques ". Departement...Valbonne, France. [Dorme 96]: R.Dorme "Nouveaux Operateurs Genetiques Appliques ä SAT" EERIE, Pare Sqientifique Georges Besse F-3000 Nimes

  1. Zirconium and technetium recovery and partitioning in the presence of actinides in modified Purex process for ATW program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dzekun, E.G.; Fedorov, Y.S.; Galkin, B.Y.; Lyubtsev, R.I.; Mashkin, A.N.; Mishin, E.N.; Zilberman, B.Y. [Radievyj Inst., Leningrad (Russian Federation)

    1994-12-31

    The modified Purex process flowsheet is based on combination of all irradiated materials, their joint dissolution and reprocessing as a NPP spent fuel solution with abnormal Pu content after addition of recycled depleted U concentrate. Some groups of long-lived radionuclides could be completely recovered and localized at the stage of extraction reprocessing using 30% TBP. Studies were conducted for 10 y to develop the process for recovery, concentration, and localization of U, Pu, Np, Tc, and Zr within 1st extraction cycle. Actinides are recovered from high-level raffinate of this cycle after evaporation and feed adjustment. Results in this report show that combined deep recovery of several elements from highly irradiated materials by TBP extraction, for further transmutation, is possible. Selective stripping of Zr from solvent phase containing U, Pu, Np, and Tc is quite effective. Development of the modified Purex process is not complete; main problem to be solved should be oxide separation from the loop and permissible storage duration before reprocessing and reuse in the loop.

  2. Topographic changes in Ni-5at.%W substrate after annealing under conditions of buffer layer crystallization

    DEFF Research Database (Denmark)

    Wulff, Anders Christian; Mishin, Oleg; Grivel, Jean-Claude

    2012-01-01

    and that the average depth of grain boundary grooves increased considerably for certain boundary types. Grooves at general high angle boundaries and Σ3 boundaries with large deviations from the ideal twin relationship were found to be more sensitive to the additional heat-treatment than grooves at low angle and true...... twin boundaries. Average groove widths increased for all boundary types. Despite the observed changes in the extent of grain boundary grooving, the mean surface roughness was almost identical before and after the additional annealing. © 2012 Published by Elsevier B.V. Selection and/or peer-review under...

  3. Nuclear knowledge-management. A core competence of VGB; Uebergreifendes Wissensmanagement fuer Kernkraftwerke. Eine VGB-Kernkompetenz

    Energy Technology Data Exchange (ETDEWEB)

    Pamme, Hartmut [RWE Power AG, Essen (Germany). Steuerung Kernkraftwerke

    2009-07-01

    It is a well established expectation that utilities/operators of nuclear power plants communicate their own operational situation and are able to comment promptly on any findings and events in the international nuclear scene. In order to gain synergies on knowledge management, utilities have been using VGB as common platform for many years. The paper describes the generic expectations concerning knowledge management towards an association like VGB. It is analysed which elements and peculiarities of modern knowledge management are already established within VGB in the nuclear field. (orig.)

  4. Source Term Analysis for the Nuclear Power Station Goesgen-Daeniken; Quelltermanalysen fuer das Kernkraftwerk Goesgen-Daeniken

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, J.P.; Megaritis, G.; Guentay, S.; Hirschmann, H.; Luebbesmeyer, D.; Lieber, K.; Jaeckel, B.; Birchley, J.; Duijvestijn, G

    2001-08-01

    Analyses are performed for three accident scenarios postulated to occur in the Goesgen Nuclear Power Plant, a 900 MWe Pressurised Water Reactor of Siemens design. The scenarios investigated comprise a Station Blackout and two separate cases of small break loss-of-coolant accident which lead, respectively, to high, intermediate and low pressure conditions in the reactor system. In each case the accident assumptions are highly pessimistic, so that the sequences span a large range of plant states and a damage phenomena. Thus the plant is evaluated for a diversity of potential safety challenges. A suite of analysis tools are used to examine the reactor coolant system response, the core heat-up, melting, fission product release from the reactor system, the transport and chemical behaviour of those fission products in the containment building, and the release of radioactivity (source term) to the environment. Comparison with reference values used by the licensing authority shows that the use of modern analysis tools and current knowledge can provide substantial reduction in the estimated source term. Of particular interest are insights gained from the analyses which indicate opportunities for operators to reduce or forestall the release. (author)

  5. The spent fuel and waste management concept of German nuclear power plants. Konzept der Entsorgung deutscher Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Nickel, H. (Forschungszentrum Juelich GmbH (Germany). Inst. fuer Reaktorwerkstoffe Technische Hochschule Aachen (Germany). Lehrstuhl fuer Reaktorwerkstoffe und Brennelemente)

    1992-07-01

    The spent fuel and waste management concept of German nuclear power plants comprises the basic legal preconditions and responsibilities, the spent fuel and radioactive waste arisings, their reprocessing and direct disposal, and the status of the Konrad, Gorleben and Morsleben repositories. Spent fuel and waste arisings also include the contaminated and activated components originating from the decommissioning of nuclear facilities. In order to close the nuclear fuel cycle, the German electricity utilities have entered into reprocessing contracts with firms in France and the United Kingdom, thereby ensuring spent fuel management up to the year 2005. All German final storage concepts provide for the emplacement of all waste, i.e. waste generating only negligible amounts of heat, in underground geologic formations. (orig.).

  6. Stade nuclear power station (KKS): four giants on tour; Kernkraftwerk Stade - KKS: Vier Riesen gehen auf Reisen

    Energy Technology Data Exchange (ETDEWEB)

    Beverungen, M.; Viermann, J. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2008-04-15

    The Stade nuclear power station was the first nuclear power plant in the Federal Republic of Germany to deliver heat in addition to electricity. Since 1984, district heat was distributed to a saltworks nearby. The power plant, which is situated on the banks of the river Elbe, was commissioned in 1972 after approximately 4 years of construction. Together with the Wuergassen plant, it was among the first commercial nuclear power plants in this country. E.ON Kernkraft holds a 2/3 interest, Vattenfall Europe a 1/3 interest in the nuclear power plant. The Stade nuclear power station was decommissioned on November 14, 2003 for economic reasons which, in part, were also politically motivated. In September 2005, the permit for demolition of the nuclear part was granted. The release from supervision under the Atomic Energy Act is expected for 2014. In the course of demolition, the 4 steam generators of the Stade nuclear power station were removed. These components, which have an aggregate weight of approx. 660 tons, are to be safely re-used in Sweden. In September 2007, the steam generators were loaded on board the Swedish special vessel, MS Sigyn, by means of a floating crane. After shipment to Sweden, heavy-duty trucks carried the components to the processing hall of Studsvik AB for further treatment. After 6 months of treatment, the contaminated inner surfaces of the tube bundles of the steam generators have been decontaminated successfully, among other items. This has increased the volume of material available for recycling and thus decreased the volume of residues. (orig.)

  7. Cutting and conditioning of the reactor pressure vessel in the NPP Wuergassen; Zerlegung und Konditionierung des Reaktordruckgefaesses im Kernkraftwerk Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Kraps, Uwe [AREVA NP GmbH, Erlangen (Germany); Duwe, Peter [E.ON Kernkraft GmbH, Bewerungen (Germany)

    2011-07-01

    NPP Wuergassen was shutdown in 1995 after 23 years of operation. Since 1997 the nuclear power plant is being dismantled. The cutting of the reactor pressure vessel internals was performed between 2003 and 2008. After decontamination the cylindrical parts of the reactor pressure vessel were dissected, the process was finalized in 2010. AREVA has now a 30 years-experience concerning repair, replacement and dismantling of reactor components. In the contribution the authors describe the process planning, manufacture and testing of appropriate remote handled tools, decontamination, dissection of the pressure vessel (320 t), conditioning, packaging and transport of the radioactive waste including radiation protection monitoring.

  8. Decommissioning of Wuergassen NPP - work on the reactor service floor; Rueckbau des Kernkraftwerkes Wuergassen - Arbeiten auf der Reaktorbedienungsebene

    Energy Technology Data Exchange (ETDEWEB)

    Ehlert, A.; Winnefeld, M. [E.ON Kernkraft GmbH, Kernkraftwerk Wuergassen, Beverungen (Germany)

    2003-07-01

    Wuergassen NPP (KWW) is located on the River Weser, some 50 km north of Kassel. It is a single unit boiling water reactor with an installed electrical generation capacity of 670 MW. The plant was constructed by AEG between 1968 and 1971 and went into operation in late 1971. Wuergassen NPP generated a total of 73 billion kWh of electricity during its period of commercial operation. Cracks were discovered in the core shroud of the reactor pressure vessel while carrying out inspection work during the annual scheduled refuelling outage in 1994. As a result of this, the plant was shut down, Various alternatives as to how to proceed further where evaluated within one year. PreussenElektra AG, the operator at the time, ultimately decided to finally shut the plant down in 1995 and to immediately begin with the decommissioning. Several independent of each other phases have been planned for the decommissioning of the nuclear power plant. The plant components are to be dismantled beginning with the non and low contaminated parts up to the more concentrated or activated installations. The whole decommissioning has been divided into a total of six phases. Theses phases are roughly depicted in fig. 1. Now that the Ministry of Economy and Small Business, Energy and Transport for the State of North Rhine Westphalia has granted permission for phases 1 to 5, all work within the framework of the direct decommissioning of Wuergassen Nuclear Power Plant can be carried out. There are at the present time approximately 500 people actively working on site. This number is made up of about 170 staff from E.ON Kernkraft GmbH and some 330 employees from outside contractors. The decommissioning tasks presented here describe the work that has been carried out in the area of the reactor service floor and specifically in the proximity of the reactor vessel. Therefore, particular requirements were necessary for the technology employed, as well as for the importance of practical radiation protection and the disposal of waste material. (orig.)

  9. Safety requirements for nuclear power plants. Content, legal validity and execution; Sicherheitsanforderungen an Kernkraftwerke. Inhalt, rechtlicher Geltungsanspruch und Vollzug

    Energy Technology Data Exchange (ETDEWEB)

    Mueller-Dehn, Christian [E.ON Kernkraft GmbH, Hannover (Germany). Nuclear Regulation and Policy

    2014-05-15

    With the approval of the 'Safety Requirements for Nuclear Power Plants' in November 2012 and the key fitting 'Interpretations' in November 2013, the decade-lasting trial on the development and actualisation of the nuclear technical regulations were successfully concluded. In terms of content the safety requirements stipulate the requirements for damage precaution as well as further safety optimisations according to paragraph 7d Atomic Energy Act (AtG). Even thought they are no international law, they tie all responsible authorities related to atomic law in the framework of existing regulations in the borders of the respective approval parameters and existing laws. In its regulations priority is given at the supervisory process to the existing approval situation and the application of safety requirements in the approval process are only acknowledged in the scope of technical modifications. (orig.)

  10. An investigation of the applicability of the new ion exchange resin, Reillex{trademark}-HPQ, in ATW separations. Milestone 4, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ashley, K.R.; Ball, J.; Grissom, M.; Williamson, M.; Cobb, S.; Young, D.; Wu, Yen-Yuan J.

    1993-09-07

    The investigations with the anion exchange resin Reillex{trademark}-HPQ is continuing along several different paths. The topics of current investigations that are reported here are: The sorption behavior of chromium(VI) on Reillex{trademark}-HPQ from nitric acid solutions and from sodium hydroxide/sodium nitrate solutions; sorption behavior of F{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; sorption behavior of Cl{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; sorption behavior of Br{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; and the Honors thesis by one of the students is attached as Appendix II (on ion exchange properties of a new macroperous resin using bromide as the model ion in aqueous nitrate solutions).

  11. Surface engineering of biaxial Gd2Zr2O7 thin films deposited on Ni–5at%W substrates by a chemical solution method

    DEFF Research Database (Denmark)

    Yue, Zhao; Grivel, Jean-Claude; Liu, Min

    2012-01-01

    backscatter diffraction. A strong dependence of the morphology and texture on the film thickness is observed, mainly due to (i) the transition of growth mode associated with the critical film thickness, i.e., increasing the film thickness leads to the grain morphology changing from 2-dimensional discs (highly...... crystal structure along the film thickness observed by a transmission electron microscope. On the basis of the enhanced understanding of the crystallization processes, we demonstrate a possibility of engineering the surface morphology and texture in the film deposited on technical substrates using...

  12. The lawfulness of a licence for the dismantling of a nuclear power plant. Zur Rechtmaessigkeit einer Genehmigung fuer den Abbau eines Kernkraftwerks

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1989-12-01

    The action was for annulment of a licence for the dismantling of the Niederaichbach nuclear power plant (KKN). The action was brought by the local government of Landshut, the argument being that the dismantling work might bring about additional release of radiation affecting the land owned by the town. The local authority had agreed in a previous hearing to the dismantling and removal of the power plant subject to certain conditions and obligations, reserving the right to make further objections, which however were not filed. The action was judged to be admissible, though without merits, as official dose assessments do not indicate a possible radiation release beyond the maximum permissible limits set by section 45 Radiation Protection Ordinance. This section requires consideration of all radiation doses possibly emanating from normal operation, and incidents or accidents in a nuclear power plant at home or abroad. The preclusive effect does not extend to facts emerging after end of the period allowed for raising objections. Any new facts emerging after that period need not be introduced in the licensing procedure by way of a motion to restore the original status, but can be asserted by way of raising objections or bringing an action in court. Regensburg Administrative Court, decision of 13 March 1989, - RN 5 K 88 1274 -. (orig./HP).

  13. Retrofitting and service of nuclear power plants in operation, including those in Eastern Europe; Nachruestungen und Service fuer laufende Kernkraftwerke einschliesslich Osteuropa

    Energy Technology Data Exchange (ETDEWEB)

    Schomer, E.

    1995-12-31

    We understand the term ``retrofitting`` to include all measures which serve to increase safety or improve efficiency which are taken after the granting of the operation authorisation required under atomic energy law. These measures also ensure that the plant is updated to conform to the state of the art, i.e. is go into shape. The term ``service`` is an umbrella term, encompassing all routine services which are intended to guarantee a high availability and maintain the value of the plant. Keywords representing important service activities are listed here: regular non-destructive testing, monitoring and diagnostic systems, materials advice, repair engineering, staff training, operational analysis. (orig./UA) [Deutsch] Wir verstehen unter dem Begriff ``Nachruestung`` alle Massnahmen, die zur Erhoehung der Sicherheit oder zur Verbesserung der Wirtschaftlichkeit nach der Erteilung der atomrechtlichen Betriebsgenehmigung dienen. Mit diesen Massnahmen wird also sichergestellt, dass die Anlage mit fortschreitendem Alter dem Stand der Technik folgt, also ertuechtigt wird. Unter dem Begriff ``Service`` werden alle routinemaessigen Dienstleistungen zusammengefasst, die der Gewaehrleistung einer hohen Verfuegbarkeit sowie der Werterhaltung der Anlage dienen. Als Schlagworte wesentlicher Service-Aktivitaeten seien hier angefuehrt: wiederkehrende zerstoerungsfreie Pruefungen, Ueberwachungs- und Diagnosetechnik, Werkstoffberatung, Reparaturtechnik, Personalschulung, Betriebsanalyse. (orig./UA)

  14. Nuclear power plants in Germany. Recent developments in off-site nuclear emergency preparedness and response; Kernkraftwerke in Deutschland. Neue Entwicklungen im anlagenexternen Notfallschutz

    Energy Technology Data Exchange (ETDEWEB)

    Gering, Florian [Bundesamt fuer Strahlenschutz, Oberschleissheim/Neuherberg (Germany). Abt. SW 2.2 Entscheidungshilfesysteme, Lageermittlung und Kommunikation

    2014-10-15

    The reactor accident in Fukushima, Japan, in 2011 triggered a thorough review of the off-site emergency preparedness and response for nuclear power plants in Germany. ''Off-site emergency preparedness and response'' includes all actions to protect the public outside the fence of a nuclear power plant. This review resulted in several changes in off-site emergency preparedness and response, which are briefly described in this article. Additionally, several recent activities are described which may influence emergency preparedness and response in the future.

  15. Implementation of a radiological emergency monitoring system for Bruce Power nuclear power plant (Canada); Implementierung eines radiologischen Umgebungsueberwachungsmesssystems fuer das Kernkraftwerk Bruce Power (Kanada)

    Energy Technology Data Exchange (ETDEWEB)

    Madaric, M. [Saphymo GmbH, Frankfurt (Germany)

    2016-07-01

    The Bruce Power nuclear power plant (BP NPP) in Ontario, Canada, is the largest nuclear generating station in the world, operating 8 nuclear reactors producing 6300 MW. In correlation with Bruce Power's safety culture, ''Safety first'' and continuous improvements are essential and substantial parts of the Bruce Power philosophy and management system. After the Fukushima nuclear accident the existing radiological emergency monitoring was analyzed and improved.

  16. Self-sustaining emergency power supply for the nuclear power plant Beznau. Project AUTANOVE; Autarke Notstrom-Versorgung fuer das Kernkraftwerk Beznau (KKB). Projekt AUTANOVE

    Energy Technology Data Exchange (ETDEWEB)

    Kaeser, Roland [Axpro AG - Kernenergie, Beznau (Switzerland). Kernkraftwerk Beznau

    2010-05-15

    The NPP Beznau is sited close to the Aare with sufficient cooling water supply so that no cooling tower is necessary. The author describes the project AUTANOVE, an self-sustaining emergency power supply for the NPP Beznau, including an evaluation of the reliability for the accidental situations fire and internal flooding, external flooding and low-water, air plane crash and safety earthquake. The new system includes two new seismic qualified, physically separated emergency diesel generators, for each unit. Deterministic and probabilistic safety analyses show further increase of the already high safety level.

  17. Ultrasonic findings in the NPP Beznau. Report on the planned further procedure of the licensee; Ultraschallbefunde des Kernkraftwerks Beznau. Stellungnahme zum geplanten weiteren Vorgehen des Betreibers

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, Simone; Pistner, Christoph

    2016-03-15

    Due to the ultrasonic findings in the reactor pressure vessel of NPP Beznau-1 in 2015 the licensee had to provide a new structural integrity analysis based on the changed material properties. The authors discuss the interpretation of the findings in Beznau with in relation to similar findings in the Belgian NPPS Doel-3 and Tihange-2. A doubtless metallurgical characterization of the irregularities in the RPV wall as not possible based on ultrasonic testing only. Destructive testing of samples from the reactor pressure vessel is also not possible since no original material bearing irregularities is available.

  18. Is it possible at all to compare nuclear power plants and wind power systems?; Ist eine ganzheitliche Vergleichanalyse eines Kern-Kraftwerkes und einer Windkraftanlage ueberhaupt moeglich?

    Energy Technology Data Exchange (ETDEWEB)

    Eliasz, J.; Biwan, A. [Technische Univ. Szczecin (Poland). Lehrstuhl fuer Waermetechnik

    2005-07-01

    At first glance, it appears impossible to compare power generation technologies that are as different in their conception as nuclear power plants and wind power systems. On the other hand, if one uses a holistic approach it may be possible. The contribution lists the preconditions that are required, e.g. parameters like the life cycle of a technology, the various stages of modelling energy and mass exchange of subsystems, and the interactions between the various branches of a power generation technology. (orig.)

  19. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  20. Consequences of changed nuclear power plant lifetimes in Germany. Scenario analyses until 2035; Auswirkungen veraenderter Laufzeiten fuer Kernkraftwerke in Deutschland. Szenarioanalysen bis zum Jahre 2035

    Energy Technology Data Exchange (ETDEWEB)

    Blesl, Markus; Bruchof, David; Fahl, Ulrich; Kober, Tom; Kuder, Ralf; Beestermoeller, Robert; Goetz, Birgit; Voss, Alfred

    2011-06-01

    The report is aimed to discuss the implications of changed NPP lifetimes in Germany on energy policy, environment, energy cost and macroeconomics. An extensive scenario analysis is used considering the effects on the German energy system in the frame of the European context. It is shown that a nuclear phase-out until 2017 is technically feasible, but needs adequate replacement options that will change the German energy system in the medium term. The study shows that the time of nuclear phase-out has no significant influence on the use of renewable energies.

  1. No nuclear power plant - now final repository? What to do with small amounts of waste?; Kein Kernkraftwerk - kein Endlager? Wohin mit wenig Abfaellen?

    Energy Technology Data Exchange (ETDEWEB)

    Feinhals, Joerg [DMT GmbH und Co. KG, Hamburg (Germany)

    2015-07-01

    Countries with nuclear power plants try to find a solution for the disposal of radioactive waste. Countries that have no nuclear power plants but produce radioactive waste in medicine, industry and research and operate research reactors have a problem: the challenging question of an appropriate disposal concept. Possibilities for such a concept are discussed in this contribution, for instance a multinational final repository, near-surface disposal of low- and medium-level radioactive wastes or a small scale disposal facility (SSDF). In any case safety analyses are required.

  2. The new digital neutron flux measuring system in Wuergassen nuclear power plant. Das neue digitale Neutronenfluss-Messsystem im Kernkraftwerk Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Pilhofer, K.H. (PreussenElektra AG, Kernkraftwerk Wuergassen (KWW), Beverungen (Germany))

    1994-06-01

    The 670 MW boiling water reactor of Wuergassen Nuclear Power Plant became critical for the first time on October 22, 1971. A very important criterion for all components is the reliability. With the dew digital neutron flow-measuring system TK250, the development of the failure rate is very positive. On the occasion of the 1993 revision, the existing 12 electronic cubicles were replaced by 4 new ones. Within only three weeks, all connections to the detectors, to the safety system, the control room, the signal system and the process calculator have been made. (orig.)

  3. Handling of radioactive materials in relation to the dismantling of the nuclear power plant Wuergassen; Behandlung von radioaktiven Stoffen im Zuge des Rueckbaus des Kernkraftwerkes Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, T. [Schmutz GmbH, Weil am Rhein (Germany); Korth, P. [E.ON Kernkraft GmbH, Hannover (Germany)

    2003-07-01

    During the decommissioning activities of the nuclear power plant Wuergassen (KWW) the decontamination of power plant components to minimise radioactive waste takes a high priority in the processing of waste material. In the illustration the operation of the two abrasive blasting units (SA 1 and SA 2) and the installations of the ''decont station'' the decontamination technology and the processing of the dismantled components is described. Examples demonstrate the success of the decontamination measures. Statements regarding the development steps made to the abrasive blasting unit 1 (SA 1) and the protective clothing used complete the illustration. (orig.)

  4. Introduction of REVK for the decommissioning of the Wuergassen nuclear power plant; Einfuehrung des ReVK beim Rueckbau des Kernkraftwerkes Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, S. [E.ON Kernkraft GmbH, Kernkraftwerk Wuergassen, Beverungen (Germany); Gruendler, D.; Haider, C. [Institut fuer Sicherheitstechnologie GmbH, Koeln (Germany)

    2003-07-01

    Since the introduction of the waste management tool ReVK, planning, controlling and supervision of decommissioning at the Wuergassen nuclear power plant is supported by an integrated IT System, which in its basic functions has already been used by EWN at Greifswald nuclear power plant. The entire planning, controlling and supervision of the decommissioning at the Wuergassen nuclear powerplant is supported by the waste management tool ReVK, which in its basic functions has already been used by EWN at Greifswald nuclear power plant. The entire planning, controlling and supervision of the decommissioning at the Wuergassen nuclear power plant is supported by the waste management tool ReVK, which in its basic functions has already been used by EWN at Greifswald nuclear power plant. Integration and design of the documentation system ReVK appeared to be a challenge for the decommissioning management aiming to ensure a frictionless progress of decommissioning the NPP Wuergassen. Most important steps in this process were the check of all working instructions for any modification of operational procedures and the integration of special parts of the existing workflow into the structure of the ReVK software. All these processes are subsequently documented from the position of ReVK users. (orig.)

  5. Recycling of scrap metal from the deconstruction of the Wuergassen nuclear power plant; Recycling von Metallschrotten aus dem Rueckbau des Kernkraftwerks Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Viermann, J. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Pollmann, E.; Sander, H. [E.ON Kernkraft GmbH, Beverungen (Germany); Krause, G. [Sina Industrieservice GmbH, Pforzheim (Germany)

    2003-07-01

    After the decision to decommission the Wuergassen nuclear power plant (KWW) was reached in 1995, deconstruction was started in April 1997. The planning provides for the power plant being completely deconstructed by 2008 with the exception of the UNS building and the transport preparation hall and the plant being released from nuclear law monitoring. Up to now (31{sup st} December 2002), approx. 9,500 Mg various materials (e.g. metal scrap, cable, insulation) have been deconstructed and disposed of. In addition, radioactive waste occurs during deconstruction and this is e.g. incinerated or compacted. Most of the deconstructed masses obtained release according to release operating plans co-ordinated with experts and authorities and the tests and measurements provided for in these plans (Paragraph 29 of the Radiation Protection Laws in 2001 [1]) and were handed over to scrap dealers or landfills. This method of disposal is not possible or is not economically justifiable for approx. 30% of the quantity deconstructed, generally dismantled parts of the plant, due to the type and extent of contamination present or activation of the material. (orig.)

  6. Brennilis. First use of industrial robots in the demolition of a French nuclear power plant; Brennilis. Erster Einsatz von Industrierobotern fuer den Rueckbau eines franzoesischen Kernkraftwerks

    Energy Technology Data Exchange (ETDEWEB)

    Bienia, Harald; Noll, Thomas [NUKEM Technologies GmbH, Alzenau (Germany)

    2012-05-15

    A share of approx. 80 % nuclear electricity makes France the country with the world's largest proportion of nuclear electricity. A considerable number of French plants were commissioned more than 30 years ago. At the present time, 58 nuclear power plants out of this population are in operation, twelve have already been decommissioned and are about to be, or are being, demolished. France thus is one of the most interesting and most dynamic countries as far as future demolition projects are concerned. Current demolition projects in France have a kind of model or pilot character for the future French demolition strategy and are under particularly close supervision and inspection by the operator, Electricite de France. One of these projects is the current demolition of the CO{sub 2}-cooled heavy water reactor (EL 4) of Brennilis in Brittanny which was decommissioned in 1985. Demolition of the reactor, its primary system and ancillary systems is handled by a Franco-German consortium composed of ONET Technologies Grands Projets, France, and NUKEM Technologies, Germany. Because of the special design features of the Brennilis reactor and the boundary conditions this created, it was not possible in many cases to transfer directly German demolition techniques. The demolition technique adopted is based on the use of remotely operated robot systems not only performing disassembly but, step by step, also building up infrastructure of their own in the reactor compartment as demolition progresses. Besides the special technical features and challenges arising in this project there are also differences in licensing regulations and cultural differences which play a major role. The report concludes with a brief summary of experience accumulated. (orig.)

  7. Methods and data of probabilistic safety analysis for nuclear power plants. Status May 2015; Methoden und Daten zur probabilistischen Sicherheitsanalyse fuer Kernkraftwerke. Stand: Mai 2015

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-09-15

    The supplement for the methodology of probabilistic safety analyses includes modifications, extensions and actualizations based on recent experiences. The chapter on personnel actions has been reorganized and adapted to the status of science and technology. Especially the possibility of decision fault identification and evaluation has been included. The chapters on floods and earthquakes are revised with respect to the actual regulatory developments and the new safety requirements. An extension of the spectra of PSA methods and data for the non-power operation has not been revised with respect to the Fukushima experiences. Based on fire experiences during power operation a new section on fire during non-power operation was included.

  8. Expertise on the Goesgen-Daeniken nuclear power plant on the granting of a licence for the construction and operation of a water storage pool for fuel assemblies at the site of the power plant; Gutachten zum Gesuch der Kernkraftwerk Goesgen-Daeniken AG um Erteilung der Bewilligung fuer den Bau und Betrieb eines Brennelement-Nasslagers auf dem Areal des Kernkraftwerks

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-04-15

    On June 26, 2002, the Goesgen-Daeniken AG nuclear power plant (KKG) delivered a request to the Swiss Federal Council for the granting of a licence for the construction and operation of a water storage pool for the on-site storage of the power plant's fuel assemblies. The present report contains the results of the examination of the request by the Federal Agency for the Safety of Nuclear Installations (HSK), to check that the projected storage pool satisfies the legal requirements from the point of view of nuclear safety and protection against radioactivity. A water storage pool already exists in the reactor building of KKG. It was conceived for a fuel cycle based on the reprocessing of the spent fuel assemblies. Its capacity is not sufficient when the spent fuel assemblies are no longer reprocessed but have to be transferred and stored in the Central Intermediate Storage Facility (ZWILAG) in Wuerenlingen because their heat production is too high. The capacity of the actual water pool allows a maximum cooling time of 5-6 years, while 7-10 years are required before transfer to ZWILAG. The projected new water storage pool has to be aircraft crash and earthquake proof, in the same way that the reactor building itself has to be. It can store a maximum of 1008 fuel assemblies. The water in the pool as well as the pool walls shield the radiation from of the fuel assemblies almost completely. Each fuel assembly is put into a square steel channel. The channel walls are lined with 6.11 mg/cm{sup 2} of the neutron absorbing nuclide B-10, which guaranties the subcriticality of the water pool even if the storage pool would be entirely filled with non-irradiated fuel assemblies with the maximal allowed enrichment or the maximal allowed content of Plutonium in case of MOX fuel assemblies, which is a very conservative assumption. The heat released by decay in the spent fuel assemblies is transferred to the pool water. Storage pool cooling is carried out by natural circulation through two cooling towers which release the heat to the environment. The cooling system is designed for a maximum cooling power of 1 MW. With this system the temperature of the pool water does not exceed 80 {sup o}C. When they are retrieved from the reactor core, the fuel assemblies are first transferred to the present water storage pool within the reactor building where they remain for at least two years. During this time, most of the short-life radioactive nuclides decay such that their contribution to the production of heat becomes negligible. In the new storage pool, the total radioactivity at full loading will amount to about 10{sup 19} Bq, i.e. one order of magnitude less than the maximal activity in the present pool. As far as the volatile radio-nuclides are concerned, all noble gases except Kr-85 and all iodine isotopes except I-129 have already decayed; as a consequence, the radiological risk in the new storage pool is much lower than in the old one. As the heating rate in the new pool is more than one order of magnitude lower than that of the present one, a possible failure in the heat release system produces only a slow increase of pool water temperature of less than 1 K per hour with the maximum heating power of 1 MW. In the first phase, it is foreseen to limit the cooling power to 0.5 MW and the number of stored fuel assemblies to 504. As the number of retrieved fuel assemblies from the reactor core is about 40 per year, the first phase will last at least 10 years. After closing of the nuclear power plant at the end of its working time and its dismantling, the storage can still work independently. After examination of the whole project for the new water storage pool, HSK concludes that under some additional conditions the concept presented can be the basis for the safe operation of the pool foreseen

  9. Expertise about the request of the nuclear power plant Leibstadt for increasing the power to 3600 MW{sub th}; Gutachten zum Gesuch des Kernkraftwerks Leibstadt um Leistungserhoehung auf 3600 MW{sub th}

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-15

    On July 31, 1992, the nuclear power plant Leibstadt AG (KKL) submitted a request for a modification of its operation license for increasing the reactor thermal power to 3600 MW. In its examination, the Federal Agency for the Safety of Nuclear Installations (HSK) investigated the effects of the power increase on reactor safety, especially on the safety criteria which can limit the power. In doing this, a distinction has to be made between normal operation, design incidents and out-of-design accidents. KKL is a boiling water reactor (BWR) with a current maximum thermal power of 3,138 MW with 648 fuel assemblies in the core. Since the start of operation in February 1984, the reactor has been continuously improved and now almost fulfils the present state-of-the-art of science and technology for BWRs. After some incidents during the early years, the plant shows a high level of availability. During the past 6 years some fuel assembly damage has cast a shadow on good operational experience, but until now the collective irradiation dose of the plant staff and the environment has remained mostly below legal limits, as well as for the release of radioactive materials to the atmosphere and to the Rhine River. Calculations of core loading with the fuel assemblies presently used at KKL have shown that the operation and safety limits of the reactor core can still be preserved with a thermal power of 3600 MW. For normal operation, no objection can be raised against the power increase. This increase, however, has to be carried out step-by-step in order to gain experience concerning plant behaviour. With the higher power rating, increased dose rates are expected on systems and components, in plant rooms and in the plant area, which also leads to increased dose rates to the staff and environment. This increase has to be estimated and, possibly, correction measures will have to be taken in order to reduce them. Especially to be monitored is the dose rate increase in the machine hall. In the case of design incidents too, all safety-relevant limits and the maximal tolerable dose rates in the environment must be respected. In the context of design incidents, the 'transitory' group also constitutes the limiting case with the higher power. For the complete judgement of the safety of a nuclear power plant it is not sufficient to estimate the effects of a design incident through deterministic methods. The evaluation of the effects of out-of-design accidents needs a probabilistic safety analysis which determines the frequency as well as the consequences of an accident. The results show that KKL represents a very small risk for the environment. In KKL the measures necessary for safe operation and protection of mankind and environment at a thermal power of 3600 MW have already been taken or will be taken shortly. According to its examination, HSK concludes that there are no safety-relevant reasons speaking against an operational license for the increased thermal power. The increase will, however, have to be carried out in 4 steps of 1 year each in order to gain operational experience

  10. Target: The green meadow. How much knowledge is needed for the dismantling of nuclear power plants?; Ziel: die Gruene Wiese. Wieviel Know-how man braucht, um ein Kernkraftwerk zurueckzubauen

    Energy Technology Data Exchange (ETDEWEB)

    Bach, Friedrich-Wilhelm; Hassel, Thomas [Unterwassertechnikum Hannover (UWTH), Hannover (Germany). Inst. fuer Werkstoffkunde

    2013-07-01

    As from the year 2022, there will no nuclear power plant exist in Germany. In the contribution under consideration two scientists from the Institute of Materials Science (Hanover, Federal Republic of Germany) report on the preparations and the necessary technical knowledge in order to dismantle the highly complex nuclear facilities and to recultivate former nuclear power plant sites.

  11. Occupational safety in the nuclear power plant. The contribution of sociology to the development of a communication tool for the elimination of hazardous situations; Arbeitssicherheit im Kernkraftwerk. Der Beitrag der Sozialpsychologie zur Entwicklung eines Kommunikationsinstrumentes fuer die Behebung von Gefaehrdungssituationen

    Energy Technology Data Exchange (ETDEWEB)

    Zedler, Christien [IAOP - Institut fuer Arbeitspsychologie, Organisation und Prozessgestaltung, Berlin (Germany); Huber, Veit [E.ON Kernkraft GmbH (Germany)

    2012-11-01

    Nuclear power plant companies make efforts to enhance the operational safety in the plant. Despite a variety of measures the number of accidents at work is still too high, esp. for external personnel. Social psychological considerations were used to develop communication tools for the elimination of hazardous situations, for instance by safety dialogues between employees. The observation of hazardous situations should trigger communication and discussion on the risk of the specific situation. In the contribution practical experiences and recommendations for the realization of a safety dialogue culture in the NPP Grafenrheinfeld are summarized and illustrated by examples.

  12. 1991 results of German nuclear power plant operation. Pt. 1. Nuclear generating units of up to 1000 MW power. Deutsche Kernkraftwerke: Betriebsergebnisse 1991. T. 1. Kernkraftwerksbloecke bis 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1992-05-01

    This first part of the survey lists the operating performance data of the eight German nuclear generating units of up to 1000 MW power. The various reactors are reviewd separately, and the operating diagram of the year 1991 presents the general survey. (UA).

  13. 1990 results of German nuclear power plant operation. Pt. 1. Nuclear generating units of up to 1000 MW power. Betriebsergebnisse der westdeutschen Kernkraftwerke 1990. T. 1. Kernkraftwerksbloecke bis 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1991-05-01

    The twenty nuclear generating units with an aggregate gross installed capacity of 22,365 MWe in operation in the Federal Republic of Germany in 1990 converted 147,243 GWh of nuclear power into electricity; this is 1.5% less than the 1989 production of 149,453 GWh. The seven units of 1000 MWe maximum rated power with an aggregate gross 5152 MWe supplied 28,746 GWh in 1990 at an average availability in terms of time of 69.9% and a net capacity utilization of 63.4%. No accident occured in the nuclear power plants in the Federal Republic of Germany in 1990 which could have endangered personnel in the plants or the population in the vicinity. The operating performance data of the nuclear generating units of up to 1000 MW power are published in Part 1 of the article as excerpts from the 1990 Annual Report by the Operating Experiences Committee (ABE Committee) of the Technical Association of Operators of Large Power Plants (VGB). Part 2 of the article, which will be published at a later date, will contain the performance data of nuclear generating units of more than 1000 MW power. (orig.).

  14. Transfer of financial obligations for the disposal of nuclear waste and decommissioning of German NPP's. Legal aspects of a trust model; Sicherstellung der finanziellen Entsorgungsvorsorge fuer die Stilllegungs- und Rueckbaukosten der deutschen Kernkraftwerke. Rechtliche Randbedingungen eines Stiftungsmodells

    Energy Technology Data Exchange (ETDEWEB)

    Schewe, Markus; Wiesendahl, Stefan [Kuemmerlein Rechtsanwaelte und Notare, Essen (Germany)

    2015-04-15

    The nuclear power plant operators have to bear the costs associated with the closure and the decommissioning of the German nuclear power plants as well as the costs for the disposal of nuclear waste. For that purpose, the operators have to build up sufficient reserves for the decommissioning phase. These reserves at the end of 2013 amounted to approximately 36 billion Euro. Changing this system is discussed very so often. Last in May 2014, a public debate started dealing with the so called trust model (''Stiftungsmodell''). The press published deliberations of several operators to transfer their entire nuclear business to the Federal Republic of Germany. Under this deliberation the current nuclear power plant operations, as well as closure obligations would be contributed to trust. Further, also the reserves should be ''transferred'' to the trust. RAG-Foundation (RAG-Stiftung) - which will assume the financial obligations in connection with Germany's closure of underground coal mining activities - sometimes is cited as a role model. The article covers elements of German trust law and atomic energy law regarding such deliberations. In trust law e.g. it can be debated whether the trust should be established under public or - as in the case of RAG-Foundation - under private law. In this context we will set out the major differences between those two options. In the public law part we will notably address issues arising from individual licensing requirements for nuclear power plants and focus on questions concerning reliability, requisite qualification and organizational structures.

  15. Use of an ultrafiltration system in the Gundremmingen nuclear power plant for the treatment of nuclear process water; Einsatz einer Ultrafiltration im Kernkraftwerk Gundremmingen zur Aufbereitung von nuklearen Prozesswaessern

    Energy Technology Data Exchange (ETDEWEB)

    Krumpholz, Udo [Kernkraftwerk Gundremmingen GmbH, Gundremmingen (Germany). Teilbereich Ueberwachung - Chemie/Entsorgung; George, Carsten [Kernkraftwerk Gundremmingen GmbH, Gundremmingen (Germany). Teilbereich Technik - Maschinentechnik; Berger, Joerg [Gruenbeck Wasseraufbereitung GmbH, Hoechstaedt a.d. Donau (Germany). Energiezentralen

    2014-07-01

    Over the years, membrane filtration systems have successfully been used in conventional water treatment systems. The use of an ultrafiltration system has proven effective in the treatment of particle contaminated process water. In 2012 an ultrafiltration system was designed, installed and commissioned for the treatment of particle contaminated backwash and transport water from the condensate polishing system in the Gundremmingen nuclear power plant, units B and C. Performance data surpass the client's requirements with respect to permeate quality, flow-rate and backwash behaviour. The technology applied has proven well. (orig.)

  16. Thin-shell wormholes in neo-Newtonian theory

    Science.gov (United States)

    Övgün, Ali; Salako, Ines G.

    2017-07-01

    In this paper, we constructed an acoustic thin-shell wormhole (ATW) under neo-Newtonian theory using the Darmois-Israel junction conditions. To determine the stability of the ATW by applying the cut-and-paste method, we found the surface density and surface pressure of the ATW under neo-Newtonian hydrodynamics just after obtaining an analog acoustic neo-Newtonian solution. We focused on the effects of the neo-Newtonian parameters by performing stability analyses using different types of fluids, such as a linear barotropic fluid (LBF), a Chaplygin fluid (CF), a logarithmic fluid (LogF) and a polytropic fluid (PF). We showed that a fluid with negative energy is required at the throat to keep the wormhole stable. The ATW can be stable if suitable values of the neo-Newtonian parameters ς, A and B are chosen.

  17. American Veterinary Medical Association

    Science.gov (United States)

    ... Brian M. Atwell Dr. Atw The dangerous dog debate November 15,2017 Breed bans are popular, but ... of companion animals January 19,2017 The AVMA House of Delegates (HOD) has approved a new policy ...

  18. Dismantling according to schedule. Shotpeening facilitates dismantling of Wuergassen nuclear power plant; Rueckbau laeuft auf Hochtouren. Trockenstrahlverfahren erleichtert Rueckbau des KKW Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Franken, M.

    1999-07-01

    Wuergassen nuclear power station was decommissioned in April 1997. Dismantling activities are preceding according to schedule and will be finished by 2010. [Deutsch] Seit April 1997 ist das Kernkraftwerk Wuergassen atomrechtlich stillgelegt. Seitdem gehen die Arbeiten der ersten Rueckbauphase zuegig voran. Bis zum Jahre 2010 wird die Demontage dauern. (orig.)

  19. Decommissioning and disposal of nuclear core parts; Abbau und Entsorgung von Kernbauteilen. Strahlenschutzmassnahmen am Beispiel der Stillegungsprojekte Gundremmingen (KRB A) und Kahl (VAK)

    Energy Technology Data Exchange (ETDEWEB)

    Duempelmann, W.; Steiner, H. [Kernkraftwerk Betriebsgesellschaft mbH, Gundremmingen (Germany); Eickelpasch, N.; Hackel, W. [Versuchsatomkraftwerk Kahl GmbH (VAK), Kahl (Germany)

    1997-12-31

    The authors describe the operational procedures, the measures for radiation protection, and the experience gained in decommissioning the shut-down nuclear power plants of Gundremmingen (KRB A) and Kahl (VAK). (orig.) [Deutsch] Die Autoren beschreiben das praktische Vorgehen, die Strahlenschutzmassnahmen und die Erfahrungen beim Abbau der stillgelegten Kernkraftwerke Gundremmingen (KRB A) und Kahl (VAK). (orig.)

  20. Oceanographic and surface meteorological data collected from station ATW20 by University of Wisconsin-Milwaukee and assembled by Great Lakes Observing System (GLOS) in the Great Lakes region from 2014-07-01 to 2017-08-31 (NODC Accession 0123639)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NCEI Accession 0123639 contains oceanographic and surface meteorological data in netCDF formatted files, which follow the Climate and Forecast metadata convention...

  1. Grohnde. Documentation of the police operation during the demonstration against the NPP Grohnde on 19.03.1977 and the evacuation of the occupied cooling tower site on 23.08.1977; Grohnde. Dokumentation der Polizeieinsaetze anlaesslich der Demonstration gegen das Kernkraftwerk Grohnde am 19.03.1977 und der Raeumung des besetzten Kuehlturmgelaendes am 23.08.1977

    Energy Technology Data Exchange (ETDEWEB)

    Stricker, Michael

    2014-07-01

    The documentation of the police operation during the demonstration against the NPP Grohnde on 16.03.1977 and the evacuation of the occupied cooling tower site on 23.08.1977 covers the following issues: involved action forces: police Niedersachsen, police Nordrhein-Westfalen, police Schleswig-Holstein, police Bremen and the Bundesgrenzschutz; concept of the police operation, provisions (lodging and board) for the police, operating resources, details of the operation sequence; post-processing of the operation; the Grohnde trials.

  2. Die Energiewerke Nord GmbH. From operator of a decommissioned Russian nuclear power plant to one of Europe's leading decommissioning companies; Die Energiewerke Nord GmbH. Der Weg vom Betreiber eines stillgelegten russischen Kernkraftwerkes zu einem fuehrenden Stilllegungsunternehmen in Europa

    Energy Technology Data Exchange (ETDEWEB)

    Philipp, Marlies [Energiewerke Nord GmbH, Rubenow (Germany)

    2011-03-15

    EWN GmbH is a state-owned company with these duties: - decommissioning and demolition of the Greifswald and Rheinsberg nuclear power stations; - safe operation of the Zwischenlager Nord interim store; - development of the 'Lubminer Heide' industrial and commercial estate. Other projects for which EWN GmbH uses its know-how: - disposal of 120 decommissioned Russian nuclear submarines in Murmansk; - decommissioning and dismantling of the Juelich, NRW, AVR experimental reactor; - demolition of nuclear plants; running the Central Decontamination Operations Department at Karlsruhe, BW. Since 2008, EWN GmbH has held 25% of the shares of Deutsche Gesellschaft zum Bau- und Betrieb von Endlagern fuer Abfallstoffe mbH (DBE), a firm building and operating nuclear repositories. (orig.)

  3. Comments on conceptual questions concerning the clearance of wastes for disposal on a dump site during the decommissioning and dismantling of the nuclear power plant Obrigheim (KWO); Stellungnahme zu konzeptionellen Fragen der Freigabe zur Beseitigung auf einer Deponie bei Stilllegung und Abbau des Kernkraftwerks Obrigheim (KWO)

    Energy Technology Data Exchange (ETDEWEB)

    Kueppers, Christian

    2015-08-03

    The comments on conceptual questions concerning the clearance of wastes for disposal on a dump site during the decommissioning and dismantling of the nuclear power plant Obrigheim (KWO) cover the following issues: fundamentals of the 10 micro-Sv concept for clearance; specific regulations for the clearance of wastes from the dismantling of KWO for disposal on a dump site; disposal concept at shutdown and dismantling of KWO; measurements and control during clearance for disposal during shutdown and dismantling of KWO; documentation and reports.

  4. The development of engineered safeguards for nuclear power plants in the political and technical environment in the Federal Republic of Germany since 1955; Die Entwicklung der Sicherheitstechnik fuer Kernkraftwerke im politischen und technischen Umfeld der Bundesrepublik Deutschland seit dem Jahr 1955

    Energy Technology Data Exchange (ETDEWEB)

    Laufs, P. [Stuttgart Univ. (Germany). Philosophische-Historische Fakultaet

    2007-01-15

    The safety of nuclear power plants is determined largely by the integrity of the internally pressurized coolant containment system. The highly radioactive materials (fission products) generated within this pressurized containment (primary system) in the reactor core during nuclear power plant operation constitute an extremely great potential hazard. Catastrophic failure of the primary system, and the release into the environment of the radioactive inventory, must be avoided at all costs. Because of the high coolant pressure and the high power density, pressurized water reactors (PWR) impose particularly strict requirements with respect to reactor safety. German nuclear power plants equipped with light water reactors enjoy the reputation of being among the safest plants in the world. This frequent statement is justified in the light of the research and development work performed jointly by industry, government agencies, science, and expert bodies between the 1960s and the 1990s. The research projects, which implied considerable financial expenditures, their internationally acknowledged results, and the resultant additional backfitting measures conducted in German nuclear power plants at many billions of expenditures, were hardly noticed by the German body politic. (orig.)

  5. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Annual report 1996; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Jahresbericht 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    There were 139 notifiable events reported from nuclear power plants in Germany, and 14 reported from research reactors. These events have been anlysed for the annual report 1996 under a variety of aspects. The results do not indicate any systematics of occurrence. None of the reported events resulted in any release of radioactivity exceeding the regulatory limits, so that there were no off-site risks involved. Among the reported events, there were two belonging to category E (prompt notification), the other 133, or 14, respectively, were at lowest scale, N, and there were none belonging to scale S. (orig./DG) [Deutsch] Im Jahr 1996 wurden aus den Kernkraftwerken der Bundesrepublik Deutschland urspruenglich insgesamt 136 und aus den Forschungsreaktoren 14 Ereignisse gemeldet. Fuer den Jahresbericht wurden diese Ereignisse nach verschiedenen Gesichtspunkten analysiert. Systematische Schwachstellen wurden dabei nicht festgestellt. Bei keinem der gemeldeten Ereignisse traten Abgaben radioaktiver Stoffe oberhalb genehmigter Grenzwerte auf. Eine Gefaehrdung von Personen und der Umgebung war in keinem Fall gegeben. Im Berichtsjahr wurden zwei Ereignisse in der Kategorie E (Eilmeldung) gemeldet. Die anderen 133 bzw. 14 Ereignisse lagen in der niedrigsten Meldekategorie N (Normalmeldung). Ereignisse der Kategorie S (Sofortmeldung) traten nicht auf. (orig./DG)

  6. 46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence / Sustainable reactor operation management - safe, efficient, valuable

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Erwin [E.ON Kernkraft GmbH, Global Unit Next Generation, Hannover (Germany)

    2015-08-15

    Summary report on the following Topical Session of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015: - Sustainable Reactor Operation Management - Safe, Efficient, Valuable (Erwin Fischer) The other Sessions of the Key Topics - ''Outstanding Know-How and Sustainable Innovations'', - ''Enhanced Safety and Operation Excellence'' and - ''Decommissioning Experience and Waste Management Solutions'' have been covered in atw 7 (2015) and will be covered in further issues of atw.

  7. Studies on the deterministic and probabilistic assessment of external effects. Deterministic investigation of the robustness of German nuclear power plants against external effects under consideration of actual findings on the events to be assumed; Untersuchungen zur deterministischen und probabilistischen Bewertung von Einwirkungen von aussen (EVA-Ereignisse). Deterministische Untersuchung der Widerstandsfaehigkeit deutscher Kernkraftwerke gegen Einwirkungen von aussen, unter Beruecksichtigung aktueller Erkenntnisse hinsichtlich der anzusetzenden Einwirkungen

    Energy Technology Data Exchange (ETDEWEB)

    Sperbeck, Silvio; Strack, Christian; Thuma, Gernot

    2013-11-15

    The aim of the analyses on natural hazards described in this report was to evaluate the advantages of innovative hazard assessment methods available today over the hazard assessment methods commonly applied for German nuclear power plant sites in the past. For each hazard under consideration (earthquake, flooding, and wind loads) it has been assessed whether the new methods provide additional insights that could call for their mandatory application in future site specific hazard assessments. If no additional insights are gained, the hitherto applied methods can be considered adequate according to today's standards. In the context of this work, no areas could be identified where the hazard assessment methods stipulated in German (nuclear) regulations are generally inadequate. These methods that are commonly applied in practice do not seem to be prone to significantly underestimate the site specific hazard. Nevertheless, some newer methods allow for more precise (reduction of uncertainties) and more comprehensive (consideration of additional hazard characteristics) hazard assessments. Therefore, depending on the hazard under consideration, it could be advisable to supplement future site specific hazard assessments by some additional analyses. As the methods for some of these additional analyses are not yet fully developed, further research will be necessary to enable these amendments.

  8. Scientific-technical cooperation with foreign (esp. Europe and INSC partner countries) nuclear regulatory authorities and their technical support organizations in the fields of nuclear safety of operating nuclear power plants and on the concept evaluation of generation 3+ plants. Final report; Wissenschaftlich-Technische Zusammenarbeit (WTZ) mit auslaendischen (insbesondere in Europa und INSC-Partnerstaaten) atomrechtlichen Behoerden und deren Sachverstaendigenorganisationen zur nuklearen Sicherheit in Betrieb befindlicher Kernkraftwerke und zur Konzeptbewertung von Generation-3+-Anlagen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Wolff, Holger

    2016-09-15

    The BMUB/BfS-Project 3614I01512 forms the frame of the GRS for the scientific-technical cooperation with Technical Support Organisations and Nuclear Regulatory Authorities in the field of nuclear safety in operating NPPs and for the concept evaluation of generation 3{sup +} plants in Europe and INSC Partner Countries. In the present final project report results are described which were gained within the project duration 15.10.2014 up to the 30.09.2016 in the following working packages: Investigations following the catastrophe of Fukushima Daiichi, Evaluation of selected National Action Plans, DBA and severe accident analyses for NPP with PWR (WWER-440, WWER-1000), cooperation with INSC partner countries on DBA, BDBA and severe accident analyses for WWER plants of generation 3{sup +} and building NRA and safety evaluation capacities and decommissioning of nuclear facilities and disposal of radioactive waste. The results are preceded by an outline on the activities related to the project management and to the planning of the bilateral work.

  9. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. First quarterly report 1997; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Vierteljahresbericht 1. Quartal 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    In the reporting period, there were 24 notifiable events reported from nuclear power plants in Germany, and one from a research reactor. The tabulated list shows all events and their event-scale categories, none involving release of radioactivity exceeding the regulatory limits or posing off-site hazards to the population or the environment. All events belong to the lowest German scale category, N, standard communication of safety significance. 23 Events, or 1, respectively, belong to the INES category 0, (no or below-scale safety significance), while one incident was classified into INES category 1 (operational anomaly, no radiological significance). (orig./DG) [Deutsch] Im I. Quartal 1997 wurden 24 meldepflichtige Ereignisse aus den Kernkraftwerken und eines aus Forschungsreaktoren der Bundesrepublik Deutschland erfasst. Die Uebersichtsliste enthaelt alle 24 Ereignisse, die in diesem Zeitraum gemeldet wurden. Abgaben radioaktiver Stoffe oberhalb genehmigter Grenzwerte traten in diesem Zeitraum nicht auf. Eine Gefaehrdung von Personen und der Umgebung war in keinem Fall gegeben. Alle meldepflichtigen Ereignisse lagen in der niedrigsten behoerdlichen Meldekategorie N (Normalmeldung). Ereignisse der behoerdlichen Meldekategorie E (Eilmeldung) und der Kategorie S (Sofortmeldung) waren nicht zu verzeichnen. 23 bzw. 1 meldepflichtiges Ereignis wurde der INES-Stufe 0 (keine oder sehr geringe sicherheitstechnische, bzw. keine radiologische Bedeutung) zugeordnet. Ein Ereignis wurde der INES-Stufe 1 (betriebliche Stoerung, keine radiologische Bedeutung) zugeordnet. (orig./DG)

  10. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Fourth quarterly report 1996; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Vierteljahresbericht 4. Quartal 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    In the 4th quarter 1996, there were 28 notifiable events reported from nuclear power plants in Germany, and 4 from research reactors, all being classified and shown in a tabulated survey. There was no event resulting in release of radioactivity exceeding the regulatory limits, and there were no off-site risks to the population or the environment. All events reported belong to the lowest category N (normal communication of safety significance), and there were non of categories E or S. All reported events belonged to INES category 0 (no or below-scale safety or radiological significance). (orig./DG) [Deutsch] Im IV. Quartal 1996 wurden 28 meldepflichtige Ereignisse aus den Kernkraftwerken und 4 aus den Forschungsreaktoren der Bundesrepublik Deutschland erfasst. Die Uebersichtsliste enthaelt alle 28 Ereignisse, die in diesem Zeitraum gemeldet wurden. Abgaben radioaktiver Stoffe oberhalb genehmigter Grenzwerte traten in diesem Zeitraum nicht auf. Eine Gefaehrdung von Personen und der Umgebung war in keinem Fall gegeben. Alle meldepflichtigen Ereignisse lagen in der niedrigsten behoerdlichen Meldekategorie N (Normalmeldung). Ereignisse der behoerdlichen Meldekategorie E (Eilmeldung) und der Kategorie S (Sofortmeldung) waren nicht zu verzeichnen. Alle meldepflichtigen Ereignisse wurden der INES-Stufe 0 (keine oder sehr geringe sicherheitstechnische, bzw. keine radiologische Bedeutung) zugeordnet. (orig./DG)

  11. Browse Title Index

    African Journals Online (AJOL)

    Items 1 - 50 of 55 ... Vol 33, No 2 (2014), Effect of Akinboye practical creativity atw ork and metaphoric thinking techniques in fostering entrepreneurial self-efficacy among ... Identifying learning characteristics of the gifted Students in the Inclusive classroom among secondary schools in Nigeria: Implications for placement, ...

  12. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key Topics / Enhanced safety and operation excellence

    Energy Technology Data Exchange (ETDEWEB)

    Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Programm Nukleare Entsorgung, Sicherheit und Strahlenforschung (NUSAFE); Baumann, Erik [AREVA GmbH, Erlangen (Germany). Radiation Protection

    2016-12-15

    Summary report on the Key Topic 'Enhanced Safety and Operation Excellence' Focus Session 'Radiation Protection' of the 47{sup th} Annual Meeting on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 have been and will be covered in further issues of atw.

  13. Nanoindentation and micro-mechanical fracture toughness of electrodeposited nanocrystalline Ni-W alloy films

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, D.E.J., E-mail: david.armstrong@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); Haseeb, A.S.M.A. [Department of Mechanical Engineering, University of Malaya, 50603 Kuala Lumpur (Malaysia); Roberts, S.G.; Wilkinson, A.J. [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); Bade, K. [Institut fuer Mikrostrukturtechnik (IMT), Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-04-30

    Nanocrystalline nickel-tungsten alloys have great potential in the fabrication of components for microelectromechanical systems. Here the fracture toughness of Ni-12.7 at.%W alloy micro-cantilever beams was investigated. Micro-cantilevers were fabricated by UV lithography and electrodeposition and notched by focused ion beam machining. Load was applied using a nanoindenter and fracture toughness was calculated from the fracture load. Fracture toughness of the Ni-12.7 at.%W was in the range of 1.49-5.14 MPa {radical}m. This is higher than the fracture toughness of Si (another important microelectromechanical systems material), but considerably lower than that of electrodeposited nickel and other nickel based alloys. - Highlights: Black-Right-Pointing-Pointer Micro-scale cantilevers manufactured by electro-deposition and focused ion beam machining. Black-Right-Pointing-Pointer Nanoindenter used to perform micro-scale fracture test on Ni-13at%W micro-cantilevers. Black-Right-Pointing-Pointer Calculation of fracture toughness of electrodeposited Ni-13at%W thin films. Black-Right-Pointing-Pointer Fracture toughness values lower than that of nanocrystalline nickel.

  14. Post-exercise hypotensive responses following an acute bout of aquatic and overground treadmill walking in people post-stroke: a pilot study.

    Science.gov (United States)

    Lai, Byron; Jeng, Brenda; Vrongistinos, Konstantinos; Jung, Taeyou

    2015-06-01

    The purpose of this study is to investigate the effects of a single-bout of aquatic treadmill walking (ATW) and overground treadmill walking (OTW) on the magnitude and duration of post-exercise ambulatory blood pressure (BP) in people post-stroke. Seven people post-stroke participated in a cross-sectional comparative study. BP was monitored for up to 9 hours after a 15-minute bout of ATW and OTW at approximately 70% of maximal oxygen consumption (VO2max), performed on separate days. Mean systolic and diastolic BP values were compared between both exercise conditions and a day without exercise (control). Three hours after OTW, mean SBP increased by 9% from pre-exercise baseline compared to a 3% decrease during the control day (P exercise compared to a 1% DBP increase of the control day (P exercise (P exercise. Also, these data suggest that ATW can elicit clinically meaningful reductions in DBP and night-time SBP. Thus, it is recommended for clinicians to consider ATW as a non-pharmaceutical means to regulate DBP and promote nighttime dipping of SBP in people post-stroke. However, caution is advised during the immediate hours after exercise, a period of possible BP inflation.

  15. 47{sup th} Annual conference on nuclear technology (AMNT 2016). Key topics / Outstanding know-how and sustainable innovations - enhanced safety and operation excellence

    Energy Technology Data Exchange (ETDEWEB)

    Raetzke, Christian [CONLAR - Consulting on Nuclear Law, Licensing and Regulation, Leipzig (Germany); Fischer, Erwin [PreussenElektra GmbH, Hannover (Germany). Management Board; Mohrbach, Ludger [VGB PowerTech e.V., Essen (Germany). Competence Center ' ' Nuclear Power Plants' '

    2016-08-15

    Summary report on the Key Topics ''Outstanding Know-How and Sustainable Innovations'' and ''Enhanced Safety and Operation Excellence'' of the 47{sup th} Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 will be covered in further issues of atw.

  16. Improving CT scan capabilities with a new trauma workflow concept: simulation of hospital logistics using different CT scanner scenarios

    NARCIS (Netherlands)

    Fung Kon Jin, P. H. P.; Dijkgraaf, M. G. W.; Alons, C. L.; van Kuijk, C.; Beenen, L. F. M.; Koole, G. M.; Goslings, J. C.

    2011-01-01

    The Amsterdam Trauma Workflow (ATW) concept includes a sliding gantry CT scanner serving two mirrored (trauma) rooms. In this study, several predefined scenarios with a varying number of CT scanners and CT locations are analyzed to identify the best performing patient flow management strategy from

  17. Results of the brugge benchmark study for flooding optimization and history matching

    NARCIS (Netherlands)

    Peters, E.; Arts, R.J.; Brouwer, G.K.; Geel, C.R.; Cullick, S.; Lorentzen, R.J.; Chen, Y.; Dunlop, K.N.B.; Vossepoel, F.C.; Xu, R.; Sarma, P.; Alhutali, A.H.; Reynolds, A.C.

    2010-01-01

    In preparation for the SPE Applied Technology Workshop (ATW) held in Brugge in June 2008, a unique benchmark project was organized to test the combined use of waterflooding-optimization and history-matching methods in a closed-loop workflow. The benchmark was organized in the form of an interactive

  18. Extended Brugge benchmark case for history matching and water flooding optimization

    NARCIS (Netherlands)

    Peters, E.; Chen, Y.; Leeuwenburgh, O.; Oliver, D.S.

    2013-01-01

    The Brugge benchmark case designed for the SPE Applied Technology Workshop (ATW) held in Brugge in June 2008 has proven to be valuable for testing and comparing methods of history matching, production optimization and closed-loop optimization by its extensive use in literature. Key features that

  19. Some basic insights into nuclear power plant decommissioning; Einige grundsaetzliche Erkenntnisse fuer die Stillegung von Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Eickelpasch, N. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH (Germany); Steiner, H. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH (Germany)

    1996-06-01

    There are 14 projects running in Germany for decommissioning of a nuclear power plant, 11 of them are performed under the responsibility of the state, and 3 are projects of industrial enterprises. The two most advanced projects are that for shutdown of unit A of the KRB Gundremmingen station, and the VAK project at Kahl (VAK experimental reactor station). Both plants are operated as subsidiaries, of the utilities RWE and Bayernwerk. The conference paper gives some basic insights obtained in the course of these two projects, covering a period of several years. The results are: The two different disposal strategies allowed by the law, i.e. ``materials recycling`` and ``ultimate disposal``, should be assessed and analysed by two separate studies. Quantities and qualities of the liquid wastes to be managed after final shutdown of a plant differ from those of the preceding phase and require specific waste management planning. It is recommended to perform a radiologic analysis of the task of decontamination of the primary loop prior to dismantling work, as shown by the activities for VAK decommissioning. (orig.) [Deutsch] In Deutschland gibt es 14 stillgelegte Kernkraftwerke, 11 davon sind staatliche Projekte, 3 kommerzielle. Die beiden am weitesten fortgeschrittenen Projekte sind der Block A des Kernkraftwerkes Grundremmingen (KRB) und das Versuchsatomkraftwerk Kahl (VAK) - beides Tochtergesellschaften des RWE und Bayernwerks. Aus der Vielzahl der Erfahrungen aus dem langjaehrigen Abbau dieser Kraftwerke sollen einige wenige grundsaetzliche Erkenntnisse aufgezeigt werden. Dies sind im einzelnen - eine insbesondere wirtschaftliche Bewertung der beiden vom Gesetz her gleichwertigen Materialwege `Wiederverwertung` und `Endlager`, - die Tatsache, dass sich nach der endgueltigen Stillegung eines Kernkraftwerkes die Menge und Qualitaet der fluessigen Abfaelle wesentlich veraendert und besondere Massnahmen erfordert, - eine strahlenschutzmaessige Bewertung der Primaerkreis

  20. In silico

    Science.gov (United States)

    Kant, Kamal; Lal, Uma Ranjan; Ghosh, Manik

    2018-01-01

    Globally, reactive oxygen species have served as an alarm predecessor toward pathogenesis of copious oxidative stress-related diseases. The researchers have turned their attention toward plant-derived herbal goods due to their promising therapeutic applications with minimal side effects. Arisaema tortuosum (Wall.) Schott (ATWS) is used in the traditional medicine since ancient years, but scientific assessments are relatively inadequate and need to be unlocked. Our aim was designed to validate the ATWS tuber and leaf extracts as an inhibitor of oxidative stress using computational approach. The reported chief chemical entities of ATWS were docked using Maestro 9.3 (Schrödinger, LLC, Cambridge, USA) tool and further ATWS extracts (tubers and leaves) were validated with 2,2'-diphenyl-1-picrylhydrazyl (DPPH), 2,2'-azino-bis (3-ethylbenzothiazoline-6-sulfonic acid) diammonium salt (ABTS), ferric-reducing ability of plasma (FRAP), and sulforhodamine B assays experimentally. In silico results showed notable binding affinity of ATWS phytoconstituents with the receptor (PDB: 3ERT). Experimentally, butanolic tuber fraction confirmed promising antioxidant potential (ABTS: IC 50 : 271.67 μg/ml; DPPH: IC 50 : 723.41 μg/ml) with a noteworthy amount of FRAP (195.96 μg/mg), total phenolic content (0.087 μg/mg), and total flavonoid content (7.5 μg/mg) while chloroform fraction (leaves) showed considerable reduction in the cell viability of MCF-7 cell line. The current findings may act as a precious tool to further unlock novel potential therapeutic agents against oxidative stress. Quercetin showed top.ranked glide score with notable binding toward 3ERT receptorAmong extracts, butanolic tubers confirmed as promising antioxidant with remarkable amount of TPC and TFCIn addition, chloroform fraction (leaves) revealed considerable decline in the cell viability of MCF-7 cell line. Abbreviations used: ATWS: Arisaema tortuosum (Wall.) Schott, DPPH: 2,2'-diphenyl-1-picrylhydrazyl

  1. Announcement of the official publication in accordance with section 15, (3), and section 17 of the Nuclear Installations Ordinance, of a licence for modifications in the Wuergassen reactor station, issued on 27 April 1990, 2nd supplement to permit no. 7/10 KWW. As of 20 July, 1990. Hinweis auf die oeffentliche Bekanntmachung gemaess Paragraph 15 Abs. 3 und Paragraph 17 der Atomrechtlichen Verfahrensverordnung ueber eine Aenderungsgenehmigung vom 27. April 1990 fuer das Kernkraftwerk Wuergassen, 2. Ergaenzung zum Bescheid Nr. 7/10 KWW. Vom 20. Juli 1990

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-07-20

    The notice refers to the licence issued for installation and operation of a system for inertisation of the reactor containment as a fire protection measure, for enhancement of the sampling system to include a further O{sub 2}-measuring equipment, and for replacement of two exhaust air flaps in the pressure suppression system. (orig./HP).

  2. Pyrochemical separations technologies envisioned for the U. S. accelerator transmutation of waste system

    Energy Technology Data Exchange (ETDEWEB)

    Laidler, J. J.

    2000-02-17

    A program has been initiated for the purpose of developing the chemical separations technologies necessary to support a large Accelerator Transmutation of Waste (ATW) system capable of dealing with the projected inventory of spent fuel from the commercial nuclear power stations in the United States. The baseline process selected combines aqueous and pyrochemical processes to enable the efficient separation of uranium, technetium, iodine, and the transuranic elements from LWR spent fuel. The diversity of processing methods was chosen for both technical and economic factors. A six-year technology evaluation and development program is foreseen, by the end of which an informed decision can be made on proceeding with demonstration of the ATW system.

  3. 2011 annual meeting on nuclear technology. Section reports. Pt. 5; Jahrestagung Kerntechnik 2011. Sektionsberichte. T. 5

    Energy Technology Data Exchange (ETDEWEB)

    Klute, Stefan [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany); Oldiges, Olaf [WAK GmbH, Eggenstein-Leopoldshafen (Germany); Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (DE). Programm Nukleare Sicherheitsforschung (NUKLEAR); Baumann, Erik [AREVA NP GmbH, Erlangen (Germany)

    2011-12-15

    Summary report on 2 out of 12 sessions of the Annual Conference on Nuclear Technology held in Berlin, 17 to 19 May 2011: - Decommissioning of nuclear installations (Session 8), and - Radiation protection (Session 11). The session: - Energy economics (Session 10), and - still not published reports on sections of other sessions will be covered in further issues of atw. Reports on the sessions: - Reactor physics and methods of calculation (Session 1), - Thermodynamics and fluid dynamics (Session 2), - Safety of nuclear installations - methods, analysis, results (Session 3), - Front end of the fuel cycle, fuel elements and core components (Session 4), - Radioactive waste management, storage (Session 5), - Operation of nuclear installations (Session 6), - New build and innovations (Session 7), - Fusion technology (Session 9), and - Education, expert knowledge, know-how-transfer (Session 12) have been covered in atw 7, 8/9, 10 and 11 (2011). (orig.)

  4. Annual meeting on nuclear technology 2013. Section report. Pt. 6

    Energy Technology Data Exchange (ETDEWEB)

    Buettner, Klaus [NUKEM Technologies GmbH, Alzenau (Germany). Dept. Process Engineering; Reimann, Peter [AREVA GmbH, Erlangen (Germany). Fuel Germany F-G; Vallentin, Roger [WTI GmbH, Juelich (Germany)

    2014-02-15

    Summary report on 2 out of 12 sessions of the Annual Conference on Nuclear Technology held in Berlin, 14 to 16 May 2013: - Radioactive waste management, Storage (Section 5), and - Decommissioning of nuclear installations (Section 8). The Sessions Reactor physics and methods of calculation (Section 1), Thermodynamics and fluid dynamics (Section 2), Safety of nuclear installations - methods, analysis, results (Section 3), Front End of the Fuel Cycle, Fuel Elements and Core Components (Section 4), Operation of nuclear installations (Section 6), New build and innovations (Section 7), and Education, Fusion technology (Section 9), Radiation protection (Section 11), and Expert knowledge, Know-how-transfer (Section 12) have been covered in atw 8/9 to 12 (2013) and 1 (2014). The other sessions (Front end of the fuel cycle, fuel elements and core components; and Energy industry and Economics) will be covered in further issues of atw. (orig.)

  5. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key Topics / Enhanced safety and operation excellence and decommissioning experience and Waste management solutions

    Energy Technology Data Exchange (ETDEWEB)

    Salnikova, Tatiana [AREVA GmbH, Erlangen (Germany); Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-10-15

    Summary report on the Key Topics ''Enhanced Safety and Operation Excellence'' and ''Decommissioning Experience and Waste Management Solutions'' of the 47{sup th} Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 have been and will be covered in further issues of atw.

  6. Data Administration and Its Role at Naval Supply Systems Headquarters.

    Science.gov (United States)

    1985-09-01

    often overlookePd i iffL-g the p1 -Anna np 1p4ase en.’Id re_;sources iare niot avai labl1e when neem ~rcod. ef feti *.ne.sAnd rel iabi 1 atw o ranv...the logistic chain, it will have to tree data design from applications. An effective data administration organization is needed to ensure the data

  7. Fabrication of Ni-5 at. %W Long Tapes with CeO2 Buffer Layer by Reel-to-Reel Method

    DEFF Research Database (Denmark)

    Ma, Lin; Tian, Hui; Yue, Zhao

    2015-01-01

    A 10-m-long homemade textured Ni-5at.%W (Ni5W) long tape with a CeO2 buffer layer has been prepared successfully by means of rolling-assisted biaxially textured substrate (RABiTS) route followed by a chemical solution deposition method in a reel-to-reel manner. Globally, the Ni5W substrate and CeO2...

  8. 2008 annual meeting on nuclear technology. Pt. 1. Section reports; JAHRESTAGUNG KERNTECHNIK 2008. T. 1. Sektionsberichte

    Energy Technology Data Exchange (ETDEWEB)

    Dagan, Ron; Sanchez Espinoza, Victor Hugo [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit; Rohde, U.; Kliem, Soeren [Forschungszentrum Rossendorf e.V. (FZR), Dresden (Germany); Faber, Wolfgang; Berlepsch, Thilo v.; Spann, Holger [E.ON Kernkraft GmbH, Hannover (Germany); Schaffrath, Andreas [TUEV Nord SysTec GmbH und Co. KG, Hamburg (Germany); Schubert, Bernd [Vattenfall Europe Nuclear Energy GmbH, Hamburg (Germany); Rieger, Udo [Vattenfall Nuclear Energy GmbH, Hamburg (Germany); Christ,, Bernhard G. [NUKEM Technologies GmbH, Alzenau (Germany); Gulden, Werner [Fusion for Energy, Barcelona (Spain); Bogusch, Edgar [AREVA NP GmbH, Erlangen (Germany)

    2008-08-15

    Summary report on these 5 - out of 11 - Sections of the Annual Conference on Nuclear Technology held in Hamburg on May 27-29, 2008: - Reactor Physics and Methods of Calculation - Thermodynamics and Fluid Dynamics - Safety of Nuclear Installations - Methods, Analysis, Results - Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage - Fusion Technology. Other Sections will be covered in reports in further issues of atw. (orig.)

  9. Low Repetition Rate Copper Vapor Laser.

    Science.gov (United States)

    1977-09-01

    V ) - V or 2Vn + V . The discharge process then repeats itself. Figure 17 shows the voltage waveform at V1. In this case, the anode starts at...w fa fa Z SB H Z O CO w a H Sou IS a OS u OS Ö z H -J o o o OS w < 3 Q [d OS M =3 o- td OS CO < O < fa OS w § Id

  10. Is the Chinese Army the Real Winner in PLA Reforms

    Science.gov (United States)

    2016-10-01

    Strategic Studies, at the National Defense University. John Chen is a Research Intern in CSCMA and a Graduate Student in the Security Studies Program at...we argue that the reforms can also be read as an effort by PLAA commanders to use new joint command and control (C2) arrange- ments to reassert the...army–hosted survival skills exercise designed to increase defense cooperation between forces from the United States, Australia, and China, September 4

  11. Requirements of modernization strategies; Anforderungen an Erneuerungsstrategien

    Energy Technology Data Exchange (ETDEWEB)

    Heinbuch, R. [Bayernwerk AG, Muenchen (Germany)

    1997-11-01

    Instrumentation and control contributed a major share to the current level of safety, economic efficiency, and availability of the German nuclear power plants. German NPPs occupy a top position in this respect at international level, but novel instrumentation and digital control technology alone will not guarantee further enhancements. Therefore, the owner/operators established carefully devised maintenance and modernization strategies in order to safeguard their NPPs top position in the long run. The German NPPs are the most thoroughly automated plants of the world. In addition to the sweeping modernization strategies recommended by the plant manufacturers, based on computer-supported control, alternative modernization strategies have been considered in the evaluation process. This approach provides for room for maneuvre, for manufacturers as well as managers responsible for risk and cost optimization, which is a major task in view of the changing regulatory framework in the electricity market. (orig./CB) [Deutsch] Die Leittechnik hat an der erreichten Sicherheit, Wirtschaftlichkeit und Verfuegbarkeit der deutschen Kernkraftwerke erheblichen Anteil. Die im weltweiten Vergleich erreichten Spitzenpositionen unserer Kernkraftwerke koennen auch durch neue digitale Leittechniksysteme nicht mehr gesteigert werden. Wesentliches Ziel der Betreiber ist deshalb, diese Positionen durch sorgfaeltige Erhaltungs- und Erneuerungsstrategien auch langfristig sicherzustellen. Die deutschen Kraftwerke sind die am umfassendsten automatisierten Anlagen der Welt. Neben den von den Herstellern empfohlenen umfassenden Erneuerungsstrategien, auf der Grundlage rechnerbasierter Leittechnik, wurden alternative Erneuerungsstrategien in die Bewertung mit einbezogen. Hierdurch wurden Handlungsspielrueme geschaffen, fuer eine groessere Herstellerunabhaengigkeit und fuer die erforderliche Risiko- und Kostenoptimierung, die vom kuenftigen Strommarkt erzwungen werden. (orig./DG)

  12. Evaluation of load case ``switch-off of the high pressure pump of the emergency core cooling system``, measures of verification and in situ-test; Einstufung des Lastfalls ``Ausfall der TH-Hochdruckeinspeisepumpe``, Massnahmen zur Verifikation bis hin zum Grossversuch

    Energy Technology Data Exchange (ETDEWEB)

    Trobitz, M.; Mattheis, A. [Kernkraftwerke Gundremmingen Betriebsgesellschaft m.b.H. (Germany); Kerkhof, K.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Gurr-Beyer, C. [Buero fuer Baudynamik, Stuttgart (Germany); Hofstoetter, P. [Technischer Ueberwachungs-Verein Rheinland e.V., Koeln (Germany)

    1998-11-01

    Within the framework of periodic safety inspection of the Gundremmingen power station (RWE-Bayernwerk - KRB II), the load collectives used for the design of safety-relevant systems and components were checked for their consistency with latest updates of the design basis. It was found that there was no analytical information or study available describing a particular process and its effects, namely switch-off of the high-pressure feedwater pump of the emergency core cooling system. The paper reports the work performed for closing the gap, including preparatory analyses, accompanying measures such as vibration measurements during plant shut-down, as well as the preparation and performance of the in-situ test. The experimental results and the comparative evaluation of calculated and experimental data are presented. (orig./CB) [Deutsch] Im Rahmen der periodischen Sicherheitsueberpruefung des Kernkraftwerkes Gundremmingen (Kernkraftwerke RWE-Bayernwerk - KRB II) wurden u.a. die Lastkollektive, die zur Auslegung sicherheitstechnisch relevanter Systeme und Komponenten herangezogen wurden, auf Aktualitaet ueberprueft. Dabei zeigte sich, dass bislang fuer eine Betriebsweise - naemlich das Abschalten der Hochdruckeinspeisepumpe des nuklearen Not- und Nachkuehlsystems (TH-HD-Pumpe) - keine analytischen Untersuchungen vorliegen. Vorbetrachtungen fuer analytische Untersuchungen, begleitende Massnahmen wie Schwingungsmessungen waehrend des Anlagenstillstandes, sowie der Versuchsaufbau und die Versuchsdurchfuehrung des Anlagenversuches werden hier dargestellt. Die Ergebnisse und der Vergleich Rechnung-Messung zum Grossversuch werden in diesem Beitrag vorgestellt. (orig.)

  13. Safety Analysis for Key Design Features of KALIMER-600 Design Concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S

    2007-02-15

    This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies.

  14. Accelerator-driven Transmutation of Waste

    Science.gov (United States)

    Venneri, Francesco

    1998-04-01

    Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the

  15. Wuergassen to be decommissioned. Cracks in the core shroud did not create a hazard; Stillegung von Wuergassen beschlossen. Risse am Kernmantel stellten keine Gefaehrdung dar

    Energy Technology Data Exchange (ETDEWEB)

    Fuchs, M. [PreussenElektra AG, Hannover (Germany); Bruns, J. [Kernkraftwerk Wuergassen, Beverungen (Germany)

    1995-12-01

    Cracks in the core shroud have been detected in several boiling water reactor plants in the United States of America, Japan, Taiwan, Switzerland and Sweden. To this date, findings of this type have been made in a total of 23 plants. Also visual inspection of the core shroud of the Wuergassen Nuclear Power Station revealed cracks during the 1994 revision. The cause was found to be intergranular stress corrosion cracking (SCC) in 1.4550 type austenitic material sensitized by heat treatment. Safety assessment indicated no impairment of functional reliability either during normal operation or under accident conditions. Economic and licensing constraints prevented repair of the core shroud, although this has been achieved successfully in the United States and in Japan. Consequently, replacement of the core shroud was prepared in the planning stage, and the feasibility of this step was demonstrated. However, the fundamental modernization of the entire plant, which would have been necessary in case of replacement of the core shroud, caused PreussenElektra to decide on decommissioning the Wuergassen plant on economic grounds. This is the first decommissioning decision about a commercial nuclear power plant in the old German federal states. Inspections of the other German boiling water reactor plants did not result in any indications of cracks in the core shroud area. (orig./GL) [Deutsch] In mehreren Siedewasserreaktoranlagen in den USA, Japan, Taiwan, der Schweiz und Schweden wurden Rissbefunde am Kernmantel festgestellt. Bisher liegen in insgesamt 23 Anlagen Befunde vor. In der Revision 1994 durchgefuehrte visuelle Inspektionen am Kernmantel des Kernkraftwerkes Wuergassen ergaben ebenfalls Rissanzeigen. Als Ursache wurde interkristalline Spannungsrisskorrosion (IKSpRK) in einem durch Waermebehandlung sensibilisierten austenitischen Material 1.4550 ermittelt. Die sicherheitstechnische Bewertung ergab keine Beeintraechtigung der Funktionstuechtigkeit sowohl im Normalbetrieb

  16. Operating experience with nuclear power stations 1994; Betriebserfahrungen mit Kernkraftwerken 1994

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1995-04-01

    Within the VGB Technical Committee `Exchange of Operating Experience` - ABE - (Nuclear Engineering), an active exchange of experience has been cultivated over a period of more than 25 years. It embraces 28 nuclear power stations in Germany, Finland, France, the Netherlands, Sweden, Switzerland and Spain. This paper reports on operating results obtaned in 1994 and also on safety-related incidents, important rehabilitation measures and annual discharge rates of radio activity. (orig.) [Deutsch] Innerhalb des VGB-Fachausschusses `Austausch von Betriebserfahrungen` - ABE - (Kerntechnik) wird seit mehr als 25 Jahren ein reger Erfahrungsaustausch gepflegt. Er schliesst 28 Kernkraftwerke in Deutschland, Finnland, Frankreich, den Niederlanden, Schweden, der Schweiz und Spanien ein. Ueber die im Jahre 1994 erzielten Betriebsergebnisse sowie ueber sicherheitsrelevante Ereignisse, wichtige Umruestmassnahmen und Jahresabgaberaten an Radioaktivitaet wird berichtet. (orig.)

  17. Operating experience with nuclear power stations 1995; Betriebserfahrungen mit Kernkraftwerken 1995

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1996-04-01

    For more than 25 years, an active exchange of experience has been cultivated within the VGB Technical Committee `Exchange of Operating Experience` - ABE - (Nuclear Engineering). It embraces 28 nuclear power stations in Germany, Finland, France, The Netherlands, Sweden, Switzerland and Spain. The paper reports operating results achieved in 1995 and also safety-related incidents, important retrofitting measures and annual emission rates of radioactivity. (orig.) [Deutsch] Innerhalb des VGB-Fachausschusses `Austausch von Betriebserfahrungen` - ABE - (Kerntechnik) wird seit mehr als 25 jahren ein reger Erfahrungsaustausch gepflegt. Er schliesst 36 Kernkraftwerke in Deutschland, Finnland, Frankreich, den Niederlanden, Schweden, der Schweiz und Spanien ein. Ueber die im Jahre 1995 erzielten Betriebsergebnisse sowie ueber sicherheitsrelevante Ereignisse, wichtige Umruestmassnahmen und Jahresabgaberaten an Radioaktivitaet wird berichtet. (orig.)

  18. Operating experience from nuclear power plants in 1996; Betriebserfahrungen mit Kernkraftwerken 1996

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1997-04-01

    Within the VGB Technical Committee `Exchange of Operating Experience` (ABE), a regular exchange of experience has been cultivated for more than 25 years. It involves 28 nuclear power plants in Germany, Finland, France, The Netherlands, Sweden, Switzerland and Spain. This paper reports operating results obtained in 1996 and safety-related incidents, important retrofit measures and annual discharge rates of radioactivity. (orig.) [Deutsch] Innerhalb des VGB-Fachausschusses `Austausch von Betriebserfahrungen` - ABE - (Kerntechnik) wird seit mehr als 25 Jahren ein reger Erfahrungsaustausch gepflegt. Er schliesst 36 Kernkraftwerke in Deutschland, Finnland, Frankreich, den Niederlanden, Schweden, der Schweiz und Spanien ein. Ueber die im Jahre 1996 erzielten Betriebsergebnisse sowie ueber sicherheitsrelevante Ereignisse, wichtige Umruestmassnahmen und Jahresabgaberaten an Radioaktivitaet wird berichtet. (orig.)

  19. A task for generations. A commission plans for the future. Pt. 2. Public participation, time required, international comparison, past conflicts; Generationenaufgabe Endlagerung. Eine Kommission plant fuer die Zukunft. T. 2. Konzept zur Oeffentlichkeitsbeteiligung, Zeitbedarf, internationaler Vergleich, Konflikte der Vergangenheit

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Bernhard [E.ON Generation GmbH, Hannover (Germany); Jaeger, Gerd [RWTH Aachen Univ. (Germany)

    2016-10-15

    German Federal and State governments have committed the political foundations for the disposal of high radioactive, heat-generating waste with the Repository Site Selection Act (StandAG). The act defines a new site selection procedure and the ''Kommission Lagerung hoch radioaktiver Abfallstoffe'' (Commission Disposal of High Radioactive Waste). The Commission should evaluate the site selection process criteria, processes and decision-making basis, evaluate the StandAG and make proposals for public participation and transparency. The commission presented its final report on 5 July 2016. atw spoke with the representatives of industry, Dr. Bernhard Fischer and Prof. Dr. Gerd Jaeger, on the commission work.

  20. A task for generations. A commission plans for the future. Pt. 1; Generationenaufgabe Endlagerung. Eine Kommission plant fuer die Zukunft. T. 1. Arbeit und Umgang in der Kommission, Entsorgungspfad, Beteiligung der Oeffentlichkeit, Entscheidungskriterien

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Bernhard [PreussenElektra GmbH, Hannover (Germany); Jaeger, Gerd [RWTH Aachen (Germany)

    2016-08-15

    German Federal and State governments have committed the political foundations for the disposal of high radioactive, heat-generating waste with the Repository Site Selection Act (StandAG). The act defines a new site selection procedure and the ''Kommission Lagerung hoch radioaktiver Abfallstoffe'' (Commission Storage of High Radioactive Waste). The Commission should evaluate the site selection process criteria, processes and decision-making basis, evaluate the StandAG and make proposals for public participation and transparency. The commission presented its final report on 5 July 2016. atw spoke with the representatives of industry, Dr. Bernhard Fischer and Prof. Dr. Gerd Jaeger, on the commission work.

  1. Effect of the Ionosphere on Radiowave Systems (Based on Ionospheric Effects Symposium)

    Science.gov (United States)

    1981-04-30

    Grossi Multipath Measurements in the Athens -Salisbury T.E.P. Link .................... 297 G. Stephanou, C. Caroubalos, and N. Corallis ji An Empirical...rbbles end irregularities in the equatorial tonosphere, J. Geopbys. Res., 82, 2650, 1?77. Ossakow, S.L., sod P.K. Ch2turvedi, Morphologial stcdies of...296 A/... il@ i - 4M.,TIPAMI EM IN THE ATWS-SALISBUIr T.E.P LINK G.Stephanou, C.Caroubalos and N.Corallis UnIversity of Athens , Departbnt of Physics

  2. Medical Risk in the Future Force Unit of Employment. Results of the Army Medical Department Transformation Workshop V

    Science.gov (United States)

    2006-01-01

    40 UEyUExUA4UA3UA2 Unit of Action UA1 Figure 2.7 shows that the casualty flow was not uniformly distributed over time. This is a reasonable... UA1 18 Medical Risk in the Future Force Unit of Employment: Results of ATW V to echelons above the UA is not completely certain. In the time beyond...3.3 Time Periods When FSTs Were at Maximum Capacity RAND TR-302-3.3 UA4 (46) UA3 (23) UA2 (18) UA1 (31) 80 100 1046040200 Time (hours) U n it o f ac

  3. On Reversible Transformations of Space Elements,

    Science.gov (United States)

    1979-09-01

    the XV in (II.10) and the st. in (11.9) have the V property. 3.3. Differentiating the relation (II.10), (111.4) - xv(YV, ) we obtain (111.5) PVp =1...V.8) the quotients f()Jin terms of the ypand atwe obtain u ( pvp ;Yv’%s& := PA - -(2,s, And all these forms vanish in the neighbourhood of B0 But now...p ,...,p , ...,p ,...,p the ~’t~ 4f 4 . weight 1 and to all other pVI the weight 0. Then the terms of the weight i occur only in the term of (VI.21

  4. BWR Anticipated Transients Without Scram Leading to Instability

    Energy Technology Data Exchange (ETDEWEB)

    Cheng L. Y.; Baek J.; Cuadra, A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor power decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).

  5. Accelerator-driven transmutation of high-level waste from the defense and commercial sectors

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, C.; Arthur, E.; Beard, C. [and others

    1996-09-01

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The major goal has been to develop accelerator transmutation of waste (ATW) system designs that will thoroughly and rapidly transmute nuclear waste, including plutonium from dismantled weapons and spent reactor fuel, while generating useful electrical power and without producing a long-lived radioactive waste stream. We have identified and quantified the unique qualities of subcritical nuclear systems and their capabilities in bringing about the complete destruction of plutonium. Although the 1191 subcritical systems involved in our most effective designs radically depart from traditional nuclear reactor concepts, they are based on extrapolations of existing technologies. Overall, care was taken to retain the highly desired features that nuclear technology has developed over the years within a conservative design envelope. We believe that the ATW systems designed in this project will enable almost complete destruction of nuclear waste (conversion to stable species) at a faster rate and without many of the safety concerns associated with the possible reactor approaches.

  6. Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ball, S.J.; Nypaver, D.J.

    1994-06-01

    An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ``operator`` involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor`s fuel or metallic components.

  7. Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    H. Yamano

    2012-01-01

    Full Text Available As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS in design extension conditions (DECs. A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

  8. Disposition of nuclear waste using subcritical accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

    1998-12-31

    Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power.

  9. Electronic properties of interfaces produced by silicon wafer hydrophilic bonding

    Energy Technology Data Exchange (ETDEWEB)

    Trushin, Maxim

    2011-07-15

    The thesis presents the results of the investigations of electronic properties and defect states of dislocation networks (DNs) in silicon produced by wafers direct bonding technique. A new insight into the understanding of their very attractive properties was succeeded due to the usage of a new, recently developed silicon wafer direct bonding technique, allowing to create regular dislocation networks with predefined dislocation types and densities. Samples for the investigations were prepared by hydrophilic bonding of p-type Si (100) wafers with same small misorientation tilt angle ({proportional_to}0.5 ), but with four different twist misorientation angles Atw (being of < , 3 , 6 and 30 , respectively), thus giving rise to the different DN microstructure on every particular sample. The main experimental approach of this work was the measurements of current and capacitance of Schottky diodes prepared on the samples which contained the dislocation network at a depth that allowed one to realize all capabilities of different methods of space charge region spectroscopy (such as CV/IV, DLTS, ITS, etc.). The key tasks for the investigations were specified as the exploration of the DN-related gap states, their variations with gradually increasing twist angle Atw, investigation of the electrical field impact on the carrier emission from the dislocation-related states, as well as the establishing of the correlation between the electrical (DLTS), optical (photoluminescence PL) and structural (TEM) properties of DNs. The most important conclusions drawn from the experimental investigations and theoretical calculations can be formulated as follows: - DLTS measurements have revealed a great difference in the electronic structure of small-angle (SA) and large-angle (LA) bonded interfaces: dominating shallow level and a set of 6-7 deep levels were found in SA-samples with Atw of 1 and 3 , whereas the prevalent deep levels - in LA-samples with Atw of 6 and 30 . The critical twist

  10. The decommissioning guide. A practical help for practice?; Der Leitfaden Stillegung. Eine Erleichterung fuer die Praxis?

    Energy Technology Data Exchange (ETDEWEB)

    Klonk, H. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany); Weil, L. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany)

    1997-07-01

    Atomkernenergie - Hauptausschuss - hat auf seiner Sitzung am 4. und 5.6.96 beschlossen, den `Leitfaden zur Stillegung von Anlagen nach Paragraph 7 Atomgesetz` als Hilfsmittel - im Sinne einer Zusammenstellung rechtlich und fachlich relevanter Aspekte - fuer die Planung und Durchfuehrung von Genehmigungsverfahren und die staatliche Aufsicht im Zusammenhang mit kerntechnischen Stillegungsvorhaben heranzuziehen. Der Text des Leitfadens wurde am 12.11.96 im `Bundesanzeiger` veroeffentlicht. Inwiefern dieser Leitfaden eine tatsaechliche Erleichterung fuer die Praxis darstellt, soll im folgenden eroertert werden. In den letzten Jahren sind erhebliche Erfolge auf dem Gebiet der Stillegung von Leistungs- und Forschungsreaktoren in der Bundesrepublik Deutschland erreicht worden. Dies betrifft die technische Durchfuehrung der Demontage- und Abbauarbeiten ebenso wie die atomrechtliche Genehmigungs- und Aufsichtspraxis: Das Kernkraftwerk Niederaichbach ist vollstaendig abgebaut und der Standort freigegeben worden. Die Stillegung und die Demontage von Gundremmingen A, des MZFR, des Kernkraftwerks VAK sowie des FR-2 und des KNK II gehen planmaessig voran. Fuer die abgeschalteten Kernkraftwerke Greifswald und Rheinsberg wurden - rechtzeitig, bevor die von den Behoerden der DDR erteilten Betriebsgenehmigungen ihre Gueltigkeit verloren - von den zustaendigen Behoerden der Laender Mecklenburg-Vorpommern und Brandenburg umfassende Stillegungsgenehmigungen nach Paragraph 7 Abs. 3. AtG erteilt; die Demontage- und Zerlegungsverfahren machen Fortschritte. Bei der Konzipierung, Diskussion und Formulierung des Leitfadens Stillegung in den Jahren 1991 bis 1996 haben Vertreter aller atomrechtlichen Landesbehoerden mitgewirkt, in deren Zustaendigkeit die vorgenannten und andere Stillegungsvorhaben fallen. Es kann deshalb davon ausgegangen werden, dass bereits in der Diskussionsphase Rueckkopplungseffekte zur und mit der Praxis stattgefunden haben. (orig.)

  11. Expectations of a local radio station in connection with an emergency situation; Erwartungen eines lokalen Radiosenders im Ereignisfall

    Energy Technology Data Exchange (ETDEWEB)

    Staerkle, C. [Radio Argovia (Switzerland)

    1997-12-31

    The aspect to be discussed here actually is not what the radio station expects to happen in the event of an emergency, but rather what the population expects the radio station to do as the fastest news medium in the area in the event of a nuclear power plant incident. The radio station has to fulfill the three standard functions of providing information, entertainment, and services for the general public. The inhabitants in the area expect the radio station to broadcast information, helpful hints, and services needed in such a situation. The radio station is obliged to do its best or utmost without hesitation, thus performing one of its very specific tasks in times of an emergency. (There are four NPPs in the broadcasting area of Radio Argovia: Leibstadt, Beznau I, Beznau II, and Goesgen). The population wants information about the incident that happened, the very actual situation and the risks and hazards involved, and expected development of the situation as well as consequences. And they want it within minutes or even seconds, at any time of the day. Information given must be suitably competent, reliable and from authentic sources. This information falling short of the population`s needs or expectations, the population will have recourse to any other help within reach.(Orig./CB) [Deutsch] Es kann bei einem Ereignisfall oder besser `Stoerfall` nicht so sehr von den Erwartungen des schnellsten Mediums gesprochen werden. Vielmehr ist von den Erwartungen der Bevoelkerung an das schnellste Medium, dem Rundfunk, im Stoerfall eines Kernkraftwerkes auszugehen. Dabei hat sich das Radio seiner drei Saeulen Information, Unterhaltung und Service Public zu bedienen. Die vom Ereignis betroffenen Hoererinnen erwarten Informationen, Hilfestellungen und Dienstleistungen jeglicher Art. Diese Aufgaben kommen dem Rundfunk zu. Dieser hat sie ohne `wenn und aber` zu erfuellen. Seine Existenzberechtigung ist vor allem in Kriesenfaellen gegeben. (Im Einzugsgebiet von Radio Argovia

  12. Transmutation of Isotopes --- Ecological and Energy Production Aspects

    Science.gov (United States)

    Gudowski, Waclaw

    2000-01-01

    This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional

  13. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

  14. Thorium nuclear fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Eom, Tae Yoon; Do, Jae Bum; Choi, Yoon Dong; Park, Kyoung Kyum; Choi, In Kyu; Lee, Jae Won; Song, Woong Sup; Kim, Heong Woo

    1998-03-01

    Since thorium produces relatively small amount of TRU elements after irradiation in the reactor, it is considered one of possible media to mix with the elements to be transmuted. Both solid and molten-salt thorium fuel cycles were investigated. Transmutation concepts being studied involved fast breeder reactor, accelerator-driven subcritical reactor, and energy amplifier with thorium. Long-lived radionuclides, especially TRU elements, could be separated from spent fuel by a pyrochemical process which is evaluated to be proliferation resistance. Pyrochemical processes of IFR, MSRE and ATW were reviewed and evaluated in detail, regarding technological feasibility, compatibility of thorium with TRU, proliferation resistance, their economy and safety. (author). 26 refs., 22 figs

  15. Sensitivity Analysis of Uncertainty Parameter based on MARS-LMR Code on SHRT-45R of EBR II

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Seok-Ju; Kang, Doo-Hyuk; Seo, Jae-Seung [System Engineering and Technology Co., Daejeon (Korea, Republic of); Bae, Sung-Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Hae-Yong [Sejong University, Seoul (Korea, Republic of)

    2016-10-15

    In order to assess the uncertainty quantification of the MARS-LMR code, the code has been improved by modifying the source code to accommodate calculation process required for uncertainty quantification. In the present study, a transient of Unprotected Loss of Flow(ULOF) is selected as typical cases of as Anticipated Transient without Scram(ATWS) which belongs to DEC category. The MARS-LMR input generation for EBR II SHRT-45R and execution works are performed by using the PAPIRUS program. The sensitivity analysis is carried out with Uncertainty Parameter of the MARS-LMR code for EBR-II SHRT-45R. Based on the results of sensitivity analysis, dominant parameters with large sensitivity to FoM are picked out. Dominant parameters selected are closely related to the development process of ULOF event.

  16. Control Rod Withdrawal Events Analyses for the Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseo; Jeong, Taekyeong; Jeong, Jaeho; Chang, Wonpyo; Lee, Seungwon; An, Sangjun; Lee, Kwilim [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    To confirm the limiting condition, based on the maximum allowable reactivity insertion of 0.3 $, three cases from the end of cycle (EOC) are selected. In addition, assuming the failure of CRSS by earthquake, additional cases is defined at beginning of cycle (BOC). When the CRW occurs, the reactor can be protected by plant protection system (PPS). In this study, PPS mechanism is sequentially studied for all initiating events. For design basis accidents (DBA), the reactor can be scrammed by reactor protection system (RPS). The first and seconds RPS signals are checked during transients. When RPS is failed, so called as anticipated transient without scram (ATWS), the reactor will be protected by diverse protection system (DPS). In this study, in order to analyze various initiating events related control rod withdrawal, four kinds of operating condition is defined. TOP events are analyzed using MARS-LMR. The influence of various plant protection system such as RPS and DPS are investigated.

  17. 2009 annual meeting on nuclear technology. Pt. 1. Section reports; JAHRESTAGUNG KERNTECHNIK 2009. T. 1. Sektionsberichte

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany); Hartmann, Miks; Hoffmann, Petra Britt [Areva NP GmbH, Erlangen (Germany); Stieglitz, Robert [Forschungszentrum Karlsruhe, Eggenstein-Leopoldshafen (Germany); Hoehne, Thomas [Forschungszentrum Dresden-Rossendorf, Dresden (Germany); Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf, Inst. fuer Sicherheitsforschung, Dresden (Germany); Hollands, Thorsten [Ruhr-Univ. Bochum (RUB), Energy Systems and Energy Economics (LEE), Bochum (Germany); Sanchez Espinoza, Victor Hugo [Forschungszentrum Karlsruhe, Inst. fuer Reaktorsicherheit, Eggenstein-Leopoldshafen (Germany); Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany); Sonnenburg, H.G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Muenchen (Germany)

    2009-08-15

    Summary report on these 3 - out of 13 - Sessions of the Annual Conference on Nuclear Technology held in Dresden on May 12 to 14, 2009: Thermodynamics and Fluid Dynamics (Session 2), Safety of Nuclear Installations - Methods, Analysis, Results (Session 3), and, Front End of the Fuel Cycle, Fuel Elements and Core Components (Session 4). The other Sessions Reactor Physics and Methods of Calculation (Session 1), Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage (Session 5), Operation of Nuclear Installations (Session 6), Decommissioning of Nuclear Installations (Session 7), Fusion Technology (Session 8), Research Reactors, Neutron Sources (Session 9), Energy Industry and Economics (Session 10), Radiation Protection (Session 11), New Build and Innovations (Session 12), and Education, Expert Knowledge, Know How Transfer (Session 13) have be covered in reports in further issues of atw. (orig.)

  18. On legal requirements for construction of high temperature reactors (HTR) in Poland

    Energy Technology Data Exchange (ETDEWEB)

    Nowacki, Tomasz R. [Ministry of Economic Development, Warsaw (Poland). Dept. for Regulatory Risk Assessment

    2017-08-15

    In the July 2016 issue of atw an article has been published on the legal obstacles to the construction of HTRs in Poland. The authors have raised a number of objections to the Polish law with the main thesis of the inability, or at least a significant impediment to the construction of such installations without significant legislative intervention. The main purpose of this text is to prove that the construction of HTRs based on the existing Polish laws and regulations is possible. In addition, the author intends to clarify the particular concerns expressed in the article regarding the particular legislation and correct improper statements and interpretations of the Polish nuclear law. The article deals only with strictly legal issues and does not take a stand on the technical feasibility and reality of ambitious plans for the construction of HTRs in Poland.

  19. Probabilistic risk assessment course documentation. Volume 1: PRA fundamentals

    Energy Technology Data Exchange (ETDEWEB)

    Breeding, R J; Leahy, T J; Young, J

    1985-08-01

    The full range of PRA topics is presented, with a special emphasis on systems analysis and PRA applications. Systems analysis topics include system modeling such as fault tree and event tree construction, failure rate data, and human Reliability. The discussion of PRA applications is centered on past and present PRA based programs, such as WASH-1400 and the Interim Reliability Evaluation Program, as well as on some of the potential future applications of PRA. The relationship of PRA to generic safety issues such as station blackout and Anticipated Transient Without Scram (ATWS) is also discussed. In addition to system modeling, the major PRA tasks of accident process analysis, and consequence analysis are presented. An explanation of the results of these activities, and the techniques by which these results are derived, forms the basis for a discussion of these topics. An additional topic which is presented in this course is the topic of PRA management, organization, and evaluation. 84 figs., 41 tabs.

  20. GRSAC Users Manual

    Energy Technology Data Exchange (ETDEWEB)

    Ball, S.J.; Nypaver, D.J.

    1999-02-01

    An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model.

  1. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  2. Preliminary safety analysis for key design features of KALIMER-600

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. B.; Chang, W. P.; Suk, S. D.; Ha, K. S.; Jeong, H. Y.; Heo, S

    2004-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents in the KALIMER design with breakeven core are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2. In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated Anticipated Transient Without Scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. They are categorized as Bounding Events (BEs) because of their low probability of occurrence. In Chapter 4, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed.The performance analysis of the KALIMER-600 containment and some evaluations for the behaviors during HCDA will be performed later.

  3. Safety analysis for key design features of KALIMER-600 design concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S

    2005-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed.

  4. Attenuation of alcohol withdrawal syndrome and blood cortisol level with forced exercise in comparison with diazepam.

    Science.gov (United States)

    Motaghinejad, Majid; Bangash, Mohammad Yasan; Motaghinejad, Ozra

    2015-01-01

    Relieving withdrawal and post-abstinence syndrome of alcoholism is one of the major strategies in the treatment of alcohol addicted patients. Diazepam, chlordiazepoxide, and topiramate are the approved medications that were used for this object. To assess the role of non-pharmacologic therapy in the management of alcohol withdrawal syndrome, we analyzed effects of forced exercise by treadmill on alcohol dependent mice as an animal model. A total of 60 adult male mice were divided into 5 groups, from which 4 groups became dependent to alcohol (2 g/kg/day) for 15 days. From day 16, treatment groups were treated by diazepam (0.5mg/kg), forced exercise, and diazepam (0.5 mg/kg) concurrent with forced exercise for two weeks; And the positive control group received same dose of alcohol (2 g/kg/day) for two weeks. The negative control group received normal saline for four weeks. Finally, on day 31, all animals were observed for withdrawal signs, and Alcohol Total Withdrawal Score (ATWS) was determined. Blood cortisol levels were measured in non-fasting situations as well. Present findings showed that ATWS significantly decrease in all treatment groups in comparison with positive control group (Pdiazepam and treated by forced exercise and Pdiazepam + forced exercise). Moreover, blood cortisol level significantly decreased in all treatment groups (P<0.001). This study suggested that forced exercise and physical activity can be useful as adjunct therapy in alcoholism and can ameliorate side effects and stress situation of withdrawal syndrome periods.

  5. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

    2000-07-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

  6. Effect of W additions on the structural and magnetic properties of Ni{sub 50}Ti{sub 50−x}W{sub x} and Ti{sub 50}Ni{sub 50−x}W{sub x} systems obtained by mechanical alloying

    Energy Technology Data Exchange (ETDEWEB)

    Jara, Angelica; Arjona, Jose David; Bautista, Pedro; Gonzalez, Gema, E-mail: gemagonz@ivic.gob.ve

    2014-12-05

    Highlights: • W additions strongly affect the magnetic and structural properties of Ni-Ti. • The saturation magnetization and magnetic remanence decreases with W addition. • W additions induces amophization of Ni-Ti. - Abstract: The effect of tungsten (W{sub x}) additions (x = 0.5, 1.0, 1.5 and 2.0 at.%), on the structural and magnetic properties of the binary systems Ni{sub 50}Ti{sub 50−x} and Ti{sub 50}Ni{sub 50−x} obtained by mechanical alloying was studied. The elementary powders were milled in a Spex 8000 horizontal mill, under N{sub 2} atmosphere, for 5 and 20 h. After 20 h of milling a homogenous microstructure was observed, particularly for small W additions. For this milling time a mixed of nanocrystalline and amorphous structure was obtained. As W concentration increases (1, 1.5 and 2 at.%), in both systems, the presence of small β-W reflections and the presence of very small peaks corresponding to the formation of an incipient new phase, identified as a NiTi(W) solid solution was observed, especially evident for 2 at.%W. The saturation magnetization and magnetic remanence decreases with the addition of W down to a minimum value at 1.5 at.%W, for both systems. The samples were characterized by SEM, EDS, XRD and magnetic measurements by VSM. The structural and magnetic behavior for both ternary alloys was very similar with the W additions.

  7. Mechanized radiation testing of austenitic pipe welds. Testing of media filled pipes and determination of the flaw depth by tomosynthesis; Mechanisierte Durchstrahlungspruefung von Rundschweissnaehten. Pruefung mediengefuellter Rohrleitungen und Tiefenlagenbestimmung durch Tomosynthese

    Energy Technology Data Exchange (ETDEWEB)

    Ewert, U.; Redmer, B. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany); Mueller, J. [COMPRA GmbH, Frechen (Germany); Trobitz, M. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH, Gundremmingen (Germany); Baranov, V.A. [Institute for Introscopy, Tomsk (Russian Federation)

    1999-08-01

    A compact detection system was built for multi-angle inspection of pipes, consisting of a high-sensitivity radiometric line scanner and an ultrasonic manipulator. Improved flaw imaging quality is achieved with this system as compared to film radiography. Measurements have been carried out on site in a nuclear power plant and in a laboratory. Better flaw imaging quality was also achieved in the testing of water-filled pipes. Non-linear tomosynthesis was applied for processing and interpretation of measured data. The system delivers considerably better images of planary materials inhomogeneitites, (such as cracks and lack-of-bond defects). (orig./CB) [Deutsch] Eine hoch empfindliche radiometrische Zeilenkamera wurde mit einem Ultraschall-Manipulator zu einem Gesamtsystem aufgebaut und fuer Mehrwinkel-Inspektionen von Rohrleitungen angewandt. Bei der Inspektion von Rundschweissnaehten an Rohren mit ca. 8... 20 mm Wanddicke wurde eine Verbesserung der Bildqualitaet im Vergleich zur Filmradiographie erreicht. Diese Messungen wurden in einem Kernkraftwerk unter Vor-Ort-Bedingungen sowie im Labor ausgefuehrt. Ein signifikantes Ansteigen der Bildqualitaet wurde auch bei der Pruefung von wassergefuellten Rohren erzielt. Methoden der nicht-linearen Tomosynthese wurden fuer die Verarbeitung und Interpretation der gemessenen Projektionsdaten genutzt. Das entwickelte System gestattet eine erhebliche Verbesserung der Anzeige von planaren Materialinhomogenitaeten (z.B. Risse und Bindefehler). (orig./DGE)

  8. Analysis of core stability measurement data of advanced 9 x 9 fuel assembly in a BWR core

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, Katsuhiro; Itami, Akira; Kubo, Yuichiro; Shakudo, Taketomi [Nuclear Fuel Industries Ltd., Tokyo (Japan); Kreuter, D.; Anegawa, Takafumi; Kitamura, Hideya; Ishikawa, Masumi

    1997-05-01

    The core stability measurements were taken during the cycle-9 startup of the 1,300 MWe BWR, Kernkraftwerk Kruemmel (KKK). The core contained advanced 9 x 9 type high burn-up design reload fuel with a higher enrichment than current 8 x 8 fuel. A design feature of the advanced 9 x 9 fuel assembly (FA) is a large square water channel for enhanced neutron moderation. The measurement data as a function of core flow and power showed almost the same stability characteristics as those of the past measurement during the cycle-3 startup of the KKK core with the 8 x 8 FA. The local power range monitors (LPRM) detected neutron flux oscillations in both core-wide in-phase and half-core out-of-phase modes. The frequency-domain stability analysis using the STAIF-PK code well reproduced the measurement result that the onset of unstable operation in KKK first occurs when about half of the reactor internal pumps are operating and the other half are stopped. The stability performance of the advanced 9 x 9 FA in the core was compared with the 8 x 8 FA by a design parameter analysis with respect to thermal-hydraulic and neutronic design. It has been demonstrated by the analysis that the stability performance of the advanced 9 x 9 FA is comparable with current 8 x 8 FA. (author)

  9. Method for microbiology control in the coolant loop of Goesgen nuclear power plant using a product based on H{sub 2}O{sub 2}-silver; Verfahren zur Kontrolle der Mikrobiologie im Kuehlwasser-Kreislauf des KKW-Goesgen mit einem Produkt auf Basis H{sub 2}O{sub 2}-Silber

    Energy Technology Data Exchange (ETDEWEB)

    Braun, D.; Goemoeri, J.

    1997-12-31

    To ensure the intended high reliability and safety of nuclear power plants, treatment of the water in all loops is of high importance, as fouling may impair heat transfer, corrosive processes may damage components and pipework, entailing long periods of shut-down for repair, and high repair cost. In the Goesgen NPP, cooling of the turbine condenser is performed by an open-cycle cooling system with a natural draft cooling tower. Conditioning of the circulating water is an essential task discussed by the paper. (orig./CB) [Deutsch] Um die angestrebte hohe Verfuegbarkeit und Betriebssicherheit von Kernkraftwerken zu gewaehrleisten, kommt der Behandlung des Wassers in allen Kreislaeufen grosse Bedeutung zu, denn Ablagerungen aller Arten koennen den Waermeuebergang behindern, Korrosion kann Komponenten sowie Rohre schaedigen und zu hohen Reparaturkosten mit langen Stillstandszeiten fuehren. Im Kernkraftwerk Goesgen wird die Kuehlung des Turbinenkondensators ueber ein offenees Rueckkuehlsystem mit Naturzugkuehlturn sichergestellt. Um die eingangs gestellten Forderungen zu erfuellen, kommt der Konditionierung des Umlaufwassers besondere Bedeutung zu. (orig.)

  10. High-resolution gamma spectroscopy with whole-body and partial-body counters. Experience, recommendations. Report; Hochaufloesende Gamma-Spektrometrie an Ganz- und Teilkoerperzaehlern. Erfahrungen, Empfehlungen. Bericht

    Energy Technology Data Exchange (ETDEWEB)

    Sahre, P. [comp.

    1997-12-01

    The application of high-resolution gamma spectroscopy with whole-body and partial-body counters shows a steadily rising upward trend over the last few years. This induced the ``Arbeitskreis Inkorporationsueberwachung`` of the association ``Fachverband fuer Strahlenschutz e.V.`` to organise a meeting for joint elaboration of a guide on recommended applications of this measuring technique, based on a review of existing experience and results. A key item on the agenda of the meeting was the comparative evaluation of the Ge semiconductor detector and the NaI solid scintillation detector. (orig./CB) [Deutsch] Der Einsatz der hochaufloesenden Gammaspektroskopie in Ganz- und Teilkoerperzaehlern hat in den letzten Jahren stetig zugenommen. Der ``Arbeitskreis Inkorporationsueberwachung`` des Fachverbandes fuer Strahlenschutz e.V. hat darum bisherige Erfahrungen zusammengetragen und Empfehlungen fuer den Einsatz dieser Messtechnik erarbeitet. Der Schwerpunkt der Tagung lag beim Vergleich von Germaniumhalbleiter- mit Natriumjodid-Festszintillationsdetektoren.Tl-Detektoren eignen sich v.a. beim Vorhandensein bekannter und zeitlich konstanter Nuklidvektoren (Kernkraftwerke, Nuklearmedizin). Bei unbekanntem bzw. variablem Nuklidvektor (Stoerfall, Unfall, Forschung) sollen Ge-Detektoren benutzt werden (orig./ABI)

  11. Recent developments in Great Britain`s electricity market - The current situation; Entwicklung des Elektrizitaetsmarktes in Grossbritanien - Ein Lagebericht

    Energy Technology Data Exchange (ETDEWEB)

    Dolben, G. [Electricity Association, London (United Kingdom)

    1998-01-26

    Next year, electricity consumers in Great Britain will be in the position to choose their electricity supplier according to their own conditions. This will mean the last and most challenging phase of the transition to a deregulated electricity market. In 1990 already, a restructurization of the electricity sector in Great Britain had installed a more competition-oriented regime along with a sweeping privatisation of the entire sector (except for long-existing power plants). The article here is arrranged in five main sections: Review of the current structure of the electricity sector, the regimes of competition in electric power generation and in electricity supply to consumers, the conditions governing electric power transmission and network operators, including legal aspects, and finally aspects of price policy. (orig./CB) [Deutsch] Erst im naechsten Jahr werden alle 26 Mio Kunden in Grossbritannien in der Lage sein, ihren Stromlieferanten selbst auszuwaehlen. Dies wird die letzte und herausfordernste Stufe des Uebergangs in einen Wettbewerbsmarkt sein. Die Umstrukturierung der Elektrizitaetswirtschaft in Grossbritannien im Jahr 1990 fuehrte zu zwei Aenderungen: Der Einfuehrung eines Wettbewerbsrahmens und der Privatisierung des gesamten Sektors (mit Ausnahme der aelteren Kernkraftwerke). Dieser Aufsatz ist in fuenf Hauptabschnitte unterteilt: Eine Uebersicht ueber die Struktur der Elektrizitaetswirtschaft, Wettbewerb in der Erzeugung, Wettbewerb in der Versorgung, das Verteilungsgeschaeft und die Regulierung und schliesslich Preisentwicklungen. (orig./RHM)

  12. Advantages and limits of preventive maintenance to ensure the quality of mechanical components; Moeglichkeiten und Grenzen der vorbeugenden Instandhaltung zur Absicherung der erforderlichen Qualitaet von mechanischen Komponenten

    Energy Technology Data Exchange (ETDEWEB)

    Schoeckle, F. [Amtec Messtechnischer Service GmbH, Lauffen (Germany); Bartonicek, J. [EnBW Kernkraft GmbH, Neckarwestheim (Germany); Waidele, H.; Kockelmann, H. [Materialpruefungsanstalt (MPA), Univ. Stuttgart (Germany)

    2004-07-01

    In accordance with their safety relevance, mechanical components of a nuclear power plant can be grouped as follows: 1. Integrity concept; 2. Preventive maintenance; 3. Failure-oriented maintenance. The contribution focuses on group 2. Specifications here comprise strength, sealing quality and function, which are to be ensured by maintenance, inspection and repair in consideration of the latest state of the art. In the case of inspections, only the effects of effective damage mechanisms are considered. Inspections may be time-based or condition-based. (orig.) [German] Die mechanischen Komponenten eines Kernkraftwerks koennen entsprechend den Anforderungen an die Qualitaet im Betrieb a) Gewaehrleistung der Qualitaet (Versagen ist nicht zulaessig), b) Qualitaet ist zu erhalten (Ausfall bzw. Versagen ist im Einzelfall zulaessig), c) keine Anforderungen, in drei Gruppen eingeteilt werden: 1. Integritaetskonzept, 2. vorbeugende Instandhaltung, 3. ausfallorientierte Instandhaltung. Die erforderliche Qualitaet der Komponenten in der Gruppe 2 ''vorbeugende Instandhaltung'' wird durch die Einhaltung von Vorgaben aus Spezifikationen erreicht, die auch den Umfang der Nachweise hinsichtlich Festigkeit, Dichtheit und Funktion beinhalten. Die Erhaltung dieser anforderungsgerechten Qualitaet im Betrieb erfolgt durch Massnahmen der Wartung, Inspektion und Instandsetzung in Verbindung mit der Verfolgung des Standes von Wissenschaft und Technik. Bei den Inspektionen werden nur die Folgen von wirksamen betrieblichen Schaedigungsmechanismen erfasst. Diese Inspektionen koennen zeit- oder zustandsorientiert durchgefuehrt werden. (orig.)

  13. Comparison calculation/experiment on the load case ``shutdown of TH high pressure pumps under consideration of fluid structure interaction``; Vergleich Rechnung/Messung zum Lastfall ``Abschaltung der TH-Hochdruckpumpen unter Beruecksichtigung der Fluid-Struktur-Wechselwirkung``

    Energy Technology Data Exchange (ETDEWEB)

    Erath, W.; Nowotny, B.; Maetz, J. [KED, Rodenbach (Germany)

    1998-11-01

    Measurements of an experiment in a pipe system with pump shutdown and valve closing have been performed in the nuclear power plant KRB II. Comparative calculations of fluid and structure including interaction show an excellent agreement with the measured results. Theory and implementation of the fluid/structure interaction and the results of the comparison are described. It turns out that the consideration of the fluid/structure interaction is mostly a significant increase of the effective structural damping. (orig.) [Deutsch] Es wurden Messungen am nuklearen Nachkuehlsystem des Kernkraftwerks Gundremmingen (KRB II) bei einem Versuche mit Pumpenabschalten und Ventilschliessen durchgefuehrt. Vergleichsrechnungen der Fluid-Strukturdynamik unter echter Beruecksichtigung der Wechselwirkung ergaben eine ausgezeichnete Uebereinstimmung der Rechnung mit den Messungen. Es werden Theorie und Implementierung der Koppelung der Fluid- und Struktur-Berechnungen sowie die Vergleiche von Messung und Rechnung beschrieben. Es ergibt sich, dass die Beruecksichtigung der Wechselwirkung notwendig ist zur genaueren Berechnung von `weichen` Rohrleitungsystemen. Eine wichtige Folge der Wechselwirkung ist meist eine deutliche Erhoehung der effektiven Strukturdaempfung. (orig.)

  14. Intercrystalline stress-corrosion cracking in Nb-stabilized austenitic steel used for core internals of a BWR; Interkristalline Spannungsrisskorrosion an Nb-stabilisiertem austenistischem Stahl in Kerneinbauten eines Siedewasserreaktors

    Energy Technology Data Exchange (ETDEWEB)

    Wachter, O. [PreussenElektra AG, Hannover (Germany); Bruns, J. [PreussenElektra AG, Hannover (Germany); Wesseling, U. [Siemens/KWU, Erlangen (Germany); Kilian, R. [Siemens/KWU, Erlangen (Germany)

    1996-06-01

    Cracks detected in pipes for hot cooling water transport in German boiling water reactors have shown: titanium and Nb-stabilized materials are susceptible to intergranular stress-corrosion cracking in a hot water environment. In the Wuergassen nuclear power plant, cracks of this type were detected in the upper and lower support ring of the core shroud, and in the reinforcing rings of the upper and lower core support grid plates. The overall length of numerous single cracks was distributed over 75 up to 90% of the circumferential surface, sometimes crack depths amounted to approx. 30 mm. The cracks are caused by chromium carbide segregations at the grain interfaces. The corrosive agent primarily is H{sub 2}O{sub 2} formed by radiolysis. [Deutsch] Rissbefunde in mit heissem Reaktorwasser beaufschlagten Rohrleitungen deutscher Siedewasserreaktoren zeigten: Mit Titan- und Nb-stabilisierte Werkstoffe koennen unter den Wasserbedingungen interkristalline Spannungsrisskorrosion (IKSpRK) erleiden. Im Kernkraftwerk Wuergassen betrafen sie den oberen und unteren Auflagering des Kernmantels sowie die Verstaerkungsringe der oberen und unteren Gitterplatte. Die Gesamtlaenge zahlreicher Einzelrisse erstreckte sich auf 75 bis 90% des Umfanges, stichprobenweise wurden Risstiefen von etwa 30 mm ermittelt. Die Rissursache lag in Chromcarbid-Ausscheidungen auf den Korngrenzen. Das Korrosionspotential wird hauptsaechlich durch das radiolytisch enstehende H{sub 2}O{sub 2} gepraegt. (orig.)

  15. The probabilistic safety analysis as seen from the point of view of the Technical Inspectorates TUeV Bavaria and Saxonia; Probabilistische Sicherheitsanalyse aus der Sicht des TUEV Bayern Sachsen

    Energy Technology Data Exchange (ETDEWEB)

    Vinzens, K. [TUEV Bayern Sachsen e.V., Muenchen (Germany); Sacher, H. [TUEV Bayern Sachsen e.V., Muenchen (Germany)

    1994-07-01

    Probabilistics safety analysis (PBA) has been developed to a useful tool for an in-depth assessment of the safety of nuclear power plant. PSA methods permit a quantification of engineered plant safety covering all parameters, including in particular also the human factors. The goal pursued with PSA is not only the numeric data describing the plant safety. The systematic approach covering all parameters offers the possibility of detecting design-basis reserves for improvement, or weak points linked to certain events, possibilities of detecting contributions of plant systems to the occurrence of risky situations or accidents that cannot be managed. (orig./HP) [Deutsch] Mittlerweile haben sich probabilistische Sicherheitsanalysen zu einem nuetzlichen Werkzeug fuer die umfassende Beurteilung der Sicherheit von Kernkraftwerken entwickelt. Die angewandte Methodik erlaubt eine Quantifizierung der Anlagensicherheit unter Einbeziehung aller Einflussgroessen, inbesondere auch menschlicher Faktoren. Die Zielsetzung von Probabilistischen Sicherheitsanalysen liegt nicht allein in der zahlenmaessigen Bestimmung der Sicherheit des Kernkraftwerks. Die systematische Vorgehensweise unter Einbeziehung aller Einflussgroessen eroeffnet die Moeglichkeit, Auslegungsreserven deutlich zu machen, Schwachstellen im Zusammenhang mit bestimmten Ereignissen aufzudecken und Beitraege systemtechnischer Einrichtungen zu nicht beherrschten Ereignissen oder Gefaehrdungszustaenden zu minimieren. (orig./HP)

  16. Insights from the IAEA extrabugetary program on the safety of WWER and RBMK nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lederman, L. [IAEA, Vienna (Austria). Div. of Nuclear Instalation Safety; Goodison, D. [Nuclear Installation Inspectorate (NII), Formby (United Kingdom); Chakraborty, S. [Federal Nuclear Safety Inspectorate (HSK), Villingen (Switzerland). Safety Research and International Affairs

    1999-06-01

    An IAEA Extrabudgetary Program commenced in 1990. The objective of the Program was to assist regulators and operators of WWER and RBMK nuclear power plants to evaluate safety aspects related to the design and operation of these plants. The program terminated in 1998 and a comprehensive final report was published by the IAEA. The results showed that despite the improvements in safety already achieved, much remains to be done at individual NPPs, particularly at the WWER and RBMK plants of the first generation. Safety improvement work for these plants is essential if they are not decommissioned in the near future. The IAEA is continuing to provide specific assistance to Member States with WWER and RBMK NPPs. A specific project on WWER and RBMK safety has already been included in the IAEA Nuclear Safety program for 1999-2000. Three ongoing regional Technical Co-operation projects are also being extended until the year 2000. An important element of this assistance is to strengthen the national regulatory authorities in the countries operating these NPPs on the basis of IAEA recommendations and good international regulatory practices. Of utmost importance is to ensure that the operating organisations draw up a safety case for each NPP based on a plant specific safety analysis and that it is reviewed and approved by the national regulatory authorities. This will allow an assessment of the overall safety impact of plant modifications. (orig.) [Deutsch] Im Jahr 1990 begann ein Zusatzprogramm zum Haushalt der IAEO (Internationale Atomenergie-Organisation). Das Ziel dieses Programms war die Unterstuetzung der Regulierer und Betreiber der Kernkraftwerke des Typs WWER und RBMK bei der Bewertung der Sicherheitsaspekte hinsichtlich der Konstruktion und des Betriebs dieser KKWs. Das Programm lief im Jahr 1998 aus und die IAEO veroeffentlichte einen umfangreichen Abschlussbericht. Die Ergebnisse zeigten, dass trotz der bereits erreichten sicherheitstechnischen Verbesserungen bei

  17. Assessment of General Atomics accelerator transmutation of waste concept based on gas-turbine-modular helium cooled reactor technology.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Taiwo, T. A.; Cahalan, J. E.; Finck, P. J.

    2001-05-08

    An assessment has been performed for an Accelerator Transmutation of Waste (ATW) concept based on the use of the high temperature gas reactor technology. The concept has been proposed by General Atomics for the ATW system. The assessment was jointly conducted at Argonne National Laboratory (ANL) and Los Alamos national laboratory to assess and to define the potential candidates for the ATW system. This report represents the assessment work performed at ANL. The concept uses recycled light water reactor (LWR)-discharge-transuranic extracted from irradiated oxide fuel in a critical and sub-critical accelerator driven gas-cooled transmuter. In this concept, the transmuter operates at 600 MWt first in the critical mode for three cycles and then operates in a subcritical accelerator-driven mode for a single cycle. The transmuter contains both thermal and fast spectrum transmutation zones. The thermal zone is fueled with the TRU oxide material in the form of coated particles, which are mixed with graphite powder, packed into cylindrical compacts, and loaded in hexagonal graphite blocks with cylindrical channels; the fast zone is fueled with TRU-oxide material in the form of coated particles without the graphite powder and the graphite blocks that has been burned in the thermal region for three critical cycles and one additional accelerator-driven cycle. The fuel loaded into the fast zone is irradiated for four additional cycles. This fuel management scheme is intended to achieve a high Pu isotopes consumption in the thermal spectrum zone, and to consume the minor actinides in the fast-spectrum zone. Monte Carlo and deterministic codes have been used to assess the system performance and to determine the feasibility of achieving high TRU consumption levels. The studies revealed the potential for high consumption of Pu-239 (97%), total Pu (71%) and total TRU (64%) in the system. The analyses confirmed the need for burnable absorber for both suppressing the initial excess

  18. Altered calibration frequency due to performance of sensor tests; Foeraendrade kalibreringsintervall paa grund av genomfoerda sensortester

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, Bengt-Goeran [GSE Power Systems AB, Nykoeping (Sweden)

    2006-02-15

    The quality of the instrument components has improved since the 1970s. Nevertheless calibration is performed with the same frequency and extent today in many reactors. The maintenance routines are still the same with the result that many of the components seldom or never need to be adjusted in connection to the calibration. The calibration can furthermore only make sure the static character of the instrument components, which means that dynamic defects like filtering and delay times are missed. The report displays results from KKM (Kernkraftwerk Muehleberg in Switzerland), where dynamic investigations of the instrument system are performed annual since 1994 in cooperation with GSE Power Systems AB. The results are stored in a database with a graphic user interface, which is an important tool for the maintenance of the instrument components. HSK (nuclear power inspectorate in Switzerland) has as a result accepted an extended calibration interval from one to two years for the instrument components, as from year 2000. Calibration results from KKM are presented in the report for the instrument systems: Jet pump flow, steam flow and external recirculation flow during the years 1993-1999. The issue indicates that calibration seldom or never results in adjustment of the components. The performance of calibration demand a lot of work, this fact is also clear with numbers in the report, executed during the regular outage of the plant. The extended calibration interval at KKM since year 2000 implies that work is transferred from the hectic outage to the calm full power operation. It is therefore the author's conclusion that the introduction of this technique at KKM, in a better way than before, reaches the goal with a dynamic and static working instrument system. The fact that the introduction resulted in a reduced demand in calibration and less work during outage is an important economic advantage.

  19. A reactor noise analysis methodology for BWR core stability evaluation: application and assessment to Leibstadt stability tests

    Energy Technology Data Exchange (ETDEWEB)

    Dokhane, A. [Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland); Dokhane, A.; Ferroukhi, H.; Zimmermann, M.A. [Paul Scherrer Institut, Lab. for Reactor Physics and Systems Behavior, CH-5232 Villigen PSI (Switzerland); Aguirre, C. [Kernkraftwerk Leibstadt, CH-5325 Leibstadt (Switzerland)

    2005-07-01

    As a first step towards establishing a best-estimate methodology for the evaluation of BWR core stability parameters, i.e. the decay ratio and resonance frequency, from measured reactor noise signals, a systematic approach has recently been developed and adopted at PSI for the analysis of the Swiss BWRs. The aim is to evaluate stability tests in a consistent manner at any operating condition in the power/flow map and for any operating cycle. This methodology principally consists in a general reactor noise analysis where as much as possible information recorded during the tests is investigated prior to determining 'core representative' stability parameters along with an associated uncertainty range. A central part in this approach is that a time series analysis of all measured neutron flux signals, rather than only one or few signals, is performed. In addition, for each signal, three different model-order optimization methods are systematically employed to take into account the sensitivity upon the model-order. The adopted methodology is then applied to the evaluation of the core stability measurements performed at the Kernkraftwerk Leibstadt nuclear power plant, Switzerland, during cycles 10, 13 and 19. A total of 28 tests are hence analyzed. This is primarily done in order to obtain a broad range of tests to serve as basis for the validation of the coupled neutronic/thermal-hydraulic codes used at PSI for BWR stability calculations. In addition, in order to assess the results obtained with the current methodology, a comparative study has been carried out with respect to results from previously developed and applied procedures. The results show a good agreement between the current method and the other methods. (authors)

  20. Power engineering. Systems for energy conversion. Compact knowledge for study and profession. 4. upd. and enl. ed.; Energietechnik. Systeme zur Energieumwandlung. Kompaktwissen fuer Studium und Beruf

    Energy Technology Data Exchange (ETDEWEB)

    Zahoransky, Richard A.

    2009-07-01

    This textbook imparts to the reader a fundamental understanding for relations of energy conversion processes. It comprises the total spectra of energy engineering, starting with fundamentals of energy process engineering via description of operating power plants (all types) to energy distribution and - storage. Main topics are sustainable energy systems from renewable energy sources. combined systems (e.g. Gas/steam turbine power plants) and plants with cogeneration (e.g. modular cogeneration plants). A new chapter Kyoto-Protocol was created as a concept of emissions-free coal-fired power plants. A new wording for deregulation of energy markets was received. Numerous texts and graphs were been revised. Chapter 18 ''Deregulation of Energy Markets'' is newly revised. Due to its didactic concepts the book directs not only to students but also everybody, who is inerested into actual questions of energy engineering. (org./GL) [German] Dieses Lehrbuch vermittelt dem Leser ein grundlegendes Verstaendnis fuer die Zusammenhaenge der Energieumwandlungsprozesse. Es umfasst die gesamte Bandbreite der Energietechnik. Die Schwerpunkte reichen von nachhaltigen, erneuerbaren Energietechniken, Kombianlagen (z.B. Gas- und Dampfturbinen-Kraftwerke) ueber Anlagen mit Kraft-Waerme-Kaelte-Kopplung bis hin zum Kyoto-Protokoll. Die 4. Auflage beinhaltet erstmals Uebungsaufgaben mit ausfuehrlichen Loesungen zu den einzelnen Kapiteln. Mehrere Kapitel sind aktualisiert. Das Kapitel 18 ''Liberalisierung der Energiemaerkte'' ist neu gefasst. Aus dem Inhalt Energietechnische Grundlagen - Dampfkraftwerke - Kernkraftwerke - Gasturbinen - Kombinationskraftwerke - Stationaere Kolbenmotoren - Brennstoffzellen - Kraft-Waerme-Kaelte-Kopplung - Wasserkraft - Solartechnik - Windenergie - Biomasse - Geothermie - Energetische Muellverwertung - Energieverteilung und -speicherung - Liberalisierung der Energiemaerkte - Kyoto-Protokoll. (orig.)

  1. Transparency and efficiency through plant operations management systems; Transparenz und Effizienz durch Betriebsfuehrungssysteme

    Energy Technology Data Exchange (ETDEWEB)

    Ladage, L. [RWE Power AG, Essen (Germany). Informationsmanagement

    2001-04-01

    Plant operations management systems, being IT application systems, provide integral support of the business processes making up plant operations management. The use of plant operations management systems improves mutually interdependent factors, such as high economic performance, high availability, and maximum safety. Since its commissioning in 1988, the Emsland nuclear power station (KKE) has been run with the IBFS plant operations management system. The work flow management system (WfMS), a module of IBFS, is described as an example of job order processing. IBFS-WfMS is to optimize all processes, thus cutting costs and ensuring that processes are run and documented reliably. Assessing the savings effect achieved through the use of IBFS-WfMS clearly reveals the savings in work/time achieved by the system. These savings are quoted as approx. 4 minutes and DM 10, respectively, per working step, which corresponds to several dozens of manyears or several million DM per annum in the KKE plant under consideration. This result can be extrapolated to other plants. (orig.) [German] Betriebsfuehrungssysteme stuetzen als EDV-Anwendungssystem integral die Geschaeftsprozesse der Kraftwerksbetriebsfuehrung. Durch den Einsatz von Betriebsfuehrungssystemen werden die in gegenseitiger Abhaengigkeit befindlichen Faktoren hohe Wirtschaftlichkeit, hohe Verfuegbarkeit und groesstmoegliche Sicherheit im Verbund gefoerdert. Im Kernkraftwerk Emsland (KKE) wird seit Inbetriebnahme der Anlage im Jahr 1988 das Betriebsfuehrungssystem IBFS eingesetzt. Am Beispiel des Workflowmanagementsystems (WfMS), einem Modul des IBFS, wird die Abwicklung von Arbeitsauftraegen dargestellt. Das IBFS-WfMS soll dabei durch Optimierung aller Prozesse sowohl kostensenkend wirken als auch sicherstellen, dass die Prozesse verlaesslich abgewickelt und dokumentiert werden. Eine Abschaetzung des Einspareffektes des IBFS-WfMS zeigt deutlich die durch das System erzielten Ersparnisse an Arbeits-/Zeitaufwand auf

  2. Biological cleaning method for radioactive waste water; Biologisches Reinigungsverfahren fuer radioaktive Abwaesser

    Energy Technology Data Exchange (ETDEWEB)

    Lasch, M.; Krumpholz, U. [Kernkraftwerke Gundremmingen Betriebsgesellschaft (Germany); Eickelpasch, N.; Schohe, W. [VAK Kahl (Germany)

    1997-12-31

    Depending on the size of power plants, the waste water consisting of laundry drains and rinsing liquids from the nuclear laundry and the wash rooms within the controlled area may vary between some hundred and some thousand cubic meters. Common practice so far for water cleaning is careful filtration/sedimentation for extraction of radioactive substances, and subsequent discharge into the draining body. If radioactivity removal is insuffient, the water is evaporated for enhancing purification. The paper describes a biological method developed at the Gundremmingen reactor station. The organic matter in the waste water is removed by bacterial biodegradation, boosted by air. The time required for waste water treatment in the collecting tanks of the power plant for removal of the washing agents is approx. 10 hours, and the resulting waste water is then filtered for radioactivity removal from the water, which in the absence of detergents is much more efficient. (orig./CB) [Deutsch] In Kernkraftwerken fallen, je nach Anlagengroesse, einige hundert bis zu einigen tausend Kubikmetern Waschlaugen und Spuelwaesser aus der nuklearen Waescherei sowie aus den Duschen innerhalb des Kontrollbereichs an. Bisherige Praxis ist es, diese Waesser einer sorgfaeltigen Filtration/Sedimentation zur Abtrennung der Radioaktivitaet zu unterwerfen, um sie dann an den Vorfluter abgeben zu koennen. Sofern die Abscheidung der Radioaktivitaet nicht befriedigend moeglich ist, koennen solche Waesser durch Verdampfung gereinigt werden. Im Kernkraftwerk Gundremmingen wurde ein biologisches Verfahren zur Reinigung von Wasch- und Duschwaessern entwickelt. Dabei werden diese Waesser unter Einblasen von Luft bakteriell von organischen Bestandteilen befreit. Behandlungszeiten von ca. 10 Stunden in den im Kraftwerk vorhandenen Sammelbehaeltern reichen aus, um die Waschmittel weitestgehend abzubauen. Wenn die Waschmittel aus der Loesung entfernt worden sind, kann die Abfiltration der radioaktiven

  3. The Kruemmel ruling and its consequences; Das Kruemmel-Urteil und die Folgen

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, H. [Hochschule fuer Verwaltungswissenschaften, Speyer (Germany). Forschungsinstitut fuer Oeffentliche Verwaltung

    1997-02-01

    In its forthcoming ruling on the main issue the Higher Administrative Court is going to confirm the gist of its ruling in the interlocutory proceedings and dismiss the action brought against the permit of a modified reactor core (GE-11 fuel elements). Should an appeal be lodged against this expected ruling on points of law, the Federal Administrative Court, in line with its points argued in its ruling of August 21, 1996, will confirm the ruling by the Higher Administrative Court. At this point, at the latest, the issue will be settled formally. In the main, the ruling by the Higher Administrative Court in the interlocutory proceedings already established a remarkable degree of clarity. It is to be hoped that this clarification by the courts of law will inject some objectivity and calm into the political debate about the Kruemmel Nuclear Power Station. (orig.) [Deutsch] Das OVG wird bei seiner anstehenden Entscheidung in der Hauptsache seinen Beschluss im einstweiligen Verfahren im Ergebnis bestaetigen und die Klage gegen die Genehmigung eines veraenderten Reaktorkerns (GE-11-BE) abweisen. Fuer den Fall der Revision gegen dieses zu erwartende Urteil, wird das Bundesverwaltungsgericht entsprechend seinen eigenen Vorgaben im Urteil vom 21.8.96 das Urteil des OVG bestaetigen. Spaetenstens dann ist die Sache formell abgeschlossen; in der Sache selbst herrscht mit dem Beschluss des OVG im einstweiligen Verfahren bereits eine erfreuliche Klarheit. Es bleibt zu hoffen, dass diese gerichtliche Klarstellung ein Stueck Sachlichkeit und Unaufgeregtheit in die politische Diskussion um das Kernkraftwerk Kruemmel bringt. Dann haetten OVG und Bundesverwaltungsgericht ihre friedenstiftende Funktion erfuellt. Das Urteil des Bundesverwaltungsgerichts wird auf die Genehmigungspraxis keine ueber die bisherige Situation hinausgehenden gravierenden Auswirkungen haben - jedenfalls dann nicht, wenn der Bundesminister fuer Umwelt, Naturschutz und Reaktorsicherheit als Bundesaufsicht die

  4. Subspace Iteration Method for Complex Eigenvalue Problems with Nonsymmetric Matrices in Aeroelastic System

    Science.gov (United States)

    Pak, Chan-gi; Lung, Shun-fat

    2009-01-01

    Modern airplane design is a multidisciplinary task which combines several disciplines such as structures, aerodynamics, flight controls, and sometimes heat transfer. Historically, analytical and experimental investigations concerning the interaction of the elastic airframe with aerodynamic and in retia loads have been conducted during the design phase to determine the existence of aeroelastic instabilities, so called flutter .With the advent and increased usage of flight control systems, there is also a likelihood of instabilities caused by the interaction of the flight control system and the aeroelastic response of the airplane, known as aeroservoelastic instabilities. An in -house code MPASES (Ref. 1), modified from PASES (Ref. 2), is a general purpose digital computer program for the analysis of the closed-loop stability problem. This program used subroutines given in the International Mathematical and Statistical Library (IMSL) (Ref. 3) to compute all of the real and/or complex conjugate pairs of eigenvalues of the Hessenberg matrix. For high fidelity configuration, these aeroelastic system matrices are large and compute all eigenvalues will be time consuming. A subspace iteration method (Ref. 4) for complex eigenvalues problems with nonsymmetric matrices has been formulated and incorporated into the modified program for aeroservoelastic stability (MPASES code). Subspace iteration method only solve for the lowest p eigenvalues and corresponding eigenvectors for aeroelastic and aeroservoelastic analysis. In general, the selection of p is ranging from 10 for wing flutter analysis to 50 for an entire aircraft flutter analysis. The application of this newly incorporated code is an experiment known as the Aerostructures Test Wing (ATW) which was designed by the National Aeronautic and Space Administration (NASA) Dryden Flight Research Center, Edwards, California to research aeroelastic instabilities. Specifically, this experiment was used to study an instability

  5. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    Energy Technology Data Exchange (ETDEWEB)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L. (Science Applications International Corp., Albuquerque, NM (USA); Sandia National Labs., Albuquerque, NM (USA))

    1989-08-01

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs.

  6. Czynniki prognostyczne progresji radiologicznej w reumatoidalnym zapaleniu stawów

    Directory of Open Access Journals (Sweden)

    Piotr Wiland

    2010-08-01

    Full Text Available Możliwość przewidywania odległych konsekwencji reumatoidalnegozapalenia stawów ma zasadnicze znaczenie dla podejmowaniawłaściwych decyzji terapeutycznych u danego chorego. Markeryprognostyczne są pomocne w identyfikacji chorych o dużym ryzykuszybkiej progresji radiologicznej. Utwierdzają one lekarzaw decyzji o rozpoczęciu intensywnego leczenia u chorych z możliwąznaczną destrukcją stawów w ciągu następnych kilku lat.W artykule zaprezentowano pilotażowy model ryzyka dla przewidywaniaszybkiej progresji radiologicznej. W celu stworzenia tegomodelu posłużono się danymi, w tym wynikami badań radiologicznychpochodzącymi z badania ASPIRE, w którym porównywanomonoterapię metotreksatem z leczeniem skojarzonym metotreksatemi infliksymabem. W tym modelu wybrano wyjściowe parametry,które są łatwe do uzyskania w rutynowej praktyce klinicznej,takie jak: stężenie białka C-reaktywnego, wartość odczynuopadania krwinek czerwonych, liczbę obrzękniętych stawów orazmiano czynnika reumatoidalnego. Ten macierzowy model ryzykamoże być przydatny w ocenie ryzyka postępującego uszkodzeniastawów, szczególnie u chorych z wczesnym zapaleniem stawów.

  7. Development of Risk Management Technology/Development of Risk-Informed Application Technology

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Joon Eon; Kim, K. Y.; Ahn, K. I.; Lee, Y. H.; Lim, H. G.; Jung, W. S.; Choi, S. Y.; Han, S. J.; Ha, J. J.; Hwang, M. J.; Park, S. Y.; Yoon, C

    2007-06-15

    This project aims at developing risk-informed application technologies to enhance the safety and economy of nuclear power plant altogether. For this, the Integrated Level 1 and 2 PSA model is developed. In addition, the fire and internal flooding PSA models are improved according to the PSA standard of U.S.A. To solve the issues of domestic PSA model, the best-estimate thermal hydraulic analyses are preformed for the ATWS and LSSB. In order to reduce the uncertainty of PSA, several new PSA technologies are developed: (1) more exact quantification of large fault tree, (2) importance measure including the effects of external PSA. As feasibility studies of Option 2 and 3, the class of 6 systems' SSC are re-classified based on the risk information and the sensitivity analyses is performed for the EDG starting time, respectively. It is also improved that the methodology to identify the vital area of NPP. The research results of this project can be used in the regulatory body and the industry projects for risk-informed applications.

  8. Recent improvements of reactor physics codes in MHI

    Science.gov (United States)

    Kosaka, Shinya; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

    2015-12-01

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO's Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  9. Radioactive wastes. From where, how much, to where?; Radioaktive Abfaelle. Woher, wieviel, wohin?

    Energy Technology Data Exchange (ETDEWEB)

    Ammann, M

    2008-09-15

    This report helps to the popularization of the Nagra's works accomplished for the management and disposal of the radioactive wastes in Switzerland. The radioactive wastes are partitioned into 3 different types: high level waste (HLW), alpha-toxic waste (ATW) and low- and intermediate-level waste (L/ILW). Most of the radioactive wastes are produced in the nuclear power plants, but also by many applications in medicine, industry and research. They have to be correctly disposed of. Mankind and environment have to be protected against them in the long term. The type and quantity of the wastes are accurately known. At the nuclear power plants as well as in the central storage pool of the Zwilag AG and in the federal interim storage facility in Wuerenlingen, there is enough storage capacity for all radioactive wastes in Switzerland. Radioactive wastes can be safely disposed of in deep geological repositories for a time period long enough that the radioactivity is reduced to natural values. Nagra has proved the feasibility of such repositories and its results were accepted by the Federal Council.

  10. A study on the regulatory approach of major technical issues

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Oh, S. H.; Kang, H. J.; Kim, G. S. [Sunmoon Univ., Asan (Korea, Republic of); Lee, S. H.; Baek, W. P.; Yang, S. H.; Jeong, Y. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

    1999-02-15

    This project is to provide the regulatory direction of 4 major technical issues for the Korean Next Generation Reactors, which are parts of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are are as follows : related to the classification and acceptance criteria of plant conditions, currently used classification and safety requirement were reviewed and regulatory direction was proposed. Among multiple failures, it is identified that SBO, TLOFW, multiple SGTR and ATWS are basically to be considered for additional requirements for advanced reactors. This study reviewed risk aspects, design consideration, and trends of safety requirements, and proposed fundamental safety requirements to be applied for KNGR. Multiple steam generator tube failure is a significant safety concern because of the possibility of release of radionuclides to the environment through containment bypass. Proposed safety requirement for this event can be categorized mainly as analysis requirement, design evaluation requirement and PSA requirement; For protection of containment failure, a reasonable safety position is necessary through and integrated review of possibility of severe accident occurrence, effects of sever accident mitigation features and cost effects of these design features. With this consideration safety requirements developed are the analysis requirement, provision of protective measures and survivability/availability of protective measures.

  11. Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yo Taik

    2001-01-01

    Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented.

  12. Review of the Technical Status on the Debris Bed Cooling Model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

    2007-09-15

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris0.

  13. Chemically deposed layer sytems for the realization of YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} band conductors; Chemisch deponierte Schichtsysteme zur Realisierung von YBa{sub 2}Cu{sub 3}O{sub 7-{delta}}-Bandleitern

    Energy Technology Data Exchange (ETDEWEB)

    Engel, Sebastian

    2009-04-30

    The aim of this thesis was to produce new buffer-layer systems for biaxially texturated Ni5at%W substrates by means of chemical processes. As very promising materials for the buffer layers CaTiO{sub 3} and SrTiO{sub 3} were chosen. The production of the single layers pursued from the organometallic prestage by means of dip coating and subsequent head treatment. During the work first the single precursor solutions were to be developed. A main component of the theses forms the understanding of the texture development during the heat treatment of precursor layers on biaxially texturated metallic substrates. Based on this the growth of thick buffer layers is studied and by means of YBCO layers, which were deposed by beans of a pulsed laser, the functionality of the synthesized buffer layers proved. A further component of this thesis formes the influence of nanoscaling precipitations in thew YBCO on its superconducting properties. The YBCO deposition pursued via a variation of the TFA process, as substrate (001)-oriented SrTiO{sub 3} monocrystals were applied.

  14. High order boron transport scheme in TRAC-BF1

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    In boiling water reactors (BWR), unlike pressurized water reactors (PWR) in which the reactivity control is accomplished through movement of the control rods and boron dilution, the importance of boron transport lies in maintaining the core integrity during ATWS-kind severe accidents in which under certain circumstances a boron injection is required. This is the reason for implementing boron transport models thermal-hydraulic codes as TRAC-BF1, RELAP5 and TRACE, bringing an improvement in the accuracy of the simulations. TRAC-BF1 code provides a best estimate analysis capability for the analysis of the full range of postulated accidents in boiling water reactors systems and related facilities. The boron transport model implemented in TRAC-BF1 code is based on a calculation according to a first order accurate upwind difference scheme. There is a need in reviewing and improving this model. Four numerical schemes that solve the boron transport model have been analyzed and compared with the analytical solution that provides the Burgers equation. The studied numerical schemes are: first order Upwind, second order Godunov, second-order modified Godunov adding physical diffusion term and a third-order QUICKEST using the ULTIMATE universal limiter (UL). The modified Godunov scheme has been implemented in TRAC-BF1 source code. The results using these new schemes are presented in this paper. (author)

  15. Metal propionate synthesis of epitaxial YBa{sub 2}Cu{sub 3}O{sub 7-x} films

    Energy Technology Data Exchange (ETDEWEB)

    Ciontea, L; Petrisor, T Jr; Petrisor, T [Technical University of Cluj, Str. C. Daicoviciu Nr. 15, 400020 Cluj-Napoca (Romania); Angrisani, A; Celentano, G; Rufoloni, A; Vannozzi, A; Augieri, A; Galuzzi, V; Mancini, A [ENEA Frascati, Via Enrico Fermi 45, 00044, Frascati, Roma (Italy)], E-mail: Lelia.Ciontea@chem.utcluj.ro

    2008-02-15

    A modified TFA-MOD method, using only barium trifluoroacetate, is presented. The yttrium and copper triflouroacetates were replaced by the alcoholic solutions of Cu and Y acetates dispersed in propionic acid. Fourier transformed infrared spectroscopy (FT-IR), thermal analyses (DTA/TG) coupled with mass spectrometry (MS) and X-ray diffraction analyses were used to study the decomposition of the precursor. The method permits the shortening of the pyrolysis time by a factor 4, with respect to conventional TFA-MOD method, due to the smaller amount of evolved hydrofluoric acid. Using this method 600 nm thick YBCO films were grown both on (100)SrTiO{sub 3} and on CeO{sub 2}/YSZ/CeO{sub 2}/Pd buffered Ni-5at.%W substrates. The as obtained films exhibit good morphological, structural and superconducting properties with T{sub c} (R=0) greater than 91K and with an out-of-plain texture of 0.24 deg. and 1.9 deg., respectively.

  16. Green tea consumption after intense taekwondo training enhances salivary defense factors and antibacterial capacity.

    Directory of Open Access Journals (Sweden)

    Shiuan-Pey Lin

    Full Text Available The aim of this study was to investigate the short-term effects of green tea consumption on selected salivary defense proteins, antibacterial capacity and anti-oxidation activity in taekwondo (TKD athletes, following intensive training. Twenty-two TKD athletes performed a 2-hr TKD training session. After training, participants ingested green tea (T, caffeine 6 mg/kg and catechins 22 mg/kg or an equal volume of water (W. Saliva samples were collected at three time points: before training (BT-T; BT-W, immediately after training (AT-T; AT-W, and 30 min after drinking green tea or water (Rec-T; Rec-W. Salivary total protein, immunoglobulin A (SIgA, lactoferrin, α-amylase activity, free radical scavenger activity (FRSA and antibacterial capacity were measured. Salivary total protein, lactoferrin, SIgA concentrations and α-amylase activity increased significantly immediately after intensive TKD training. After tea drinking and 30 min rest, α-amylase activity and the ratio of α-amylase to total protein were significantly higher than before and after training. In addition, salivary antibacterial capacity was not affected by intense training, but green tea consumption after training enhanced salivary antibacterial capacity. Additionally, we observed that salivary FRSA was markedly suppressed immediately after training and quickly returned to pre-exercise values, regardless of which fluid was consumed. Our results show that green tea consumption significantly enhances the activity of α-amylase and salivary antibacterial capacity.

  17. Death qualification and prejudice: the effect of implicit racism, sexism, and homophobia on capital defendants' right to due process.

    Science.gov (United States)

    Butler, Brooke

    2007-01-01

    Two hundred venirepersons from the 12th Judicial Circuit in Bradenton, Florida completed the following measures: (1) one question that measured their level of support for the death penalty; (2) one question that categorized their death-qualification status; (3) 23 questions that measured their attitudes toward the death penalty (ATDP); (4) 22 questions that assessed their attitudes toward women (ATW); (5) 25 questions that measured their level of homophobia (H); (6) seven questions that assessed their level of modern racism (MR); (7) eight questions that measured their level of modern sexism (MS); and (8) standard demographic questions. Results indicated that as death-penalty support increased participants exhibited more positive attitudes toward the death penalty, more negative attitudes toward women, and higher levels of homophobia, modern racism, and modern sexism. Findings also suggested that death-qualified venirepersons exhibited more positive attitudes toward the death penalty and higher levels of homophobia, modern racism, and modern sexism. Finally, more positive attitudes toward the death penalty were correlated with more negative attitudes toward women and higher levels of homophobia, modern racism, and modern sexism. Legal implications are discussed. Copyright (c) 2007 John Wiley & Sons, Ltd.

  18. The German act on the reorganisation of responsibility in nuclear waste management; Des Gesetz zur Neuordnung der Verantwortung in der kerntechnischen Entsorgung

    Energy Technology Data Exchange (ETDEWEB)

    Raetzke, Christian [CONLAR Consulting on Nuclear Law and Regulation, Leipzig (Germany)

    2017-04-15

    The author discussed the Draft on the Act in the Reorganisation of Responsibility in Nuclear Waste Management in atw 12 (2016). Now, amendments are discussed, which resulted from the legislative procedure until today's draft. Significant additions affect the authorisation for the conclusion of a public-law contract between the Federal Government and the nuclear power plant operators, the deadline for the payment of the basic amount, and the option for the operation of the interim storage facilities for a transitional period by the operators on behalf of the federal company. Since the adoption of the draft act, it has become clear that the nuclear power plant operators will pay the risk premium. This will fulfil the full logic of the new system. It has also become known, that the public law contract is now ready for signing. According to the author, the act will bring a final arrangement for financing nuclear waste disposal. However, adjustment can not be avoided in practice. The concrete implementation will be a exciting topic in many ways.

  19. Burnable Absorbers with Enriched Er-167 in PWR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choe, Jiwon; Kong, Chidong; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Many advanced PWRs are required to have a 24-month operating cycle to improve plant economy, and to keep the boron concentration low to allow an adequately negative moderator feedback during any ATWS event through 100% core life. Unfortunately, longer cycles require higher uranium-235 enrichment and initial boron concentration in the reactor coolant. The amount of soluble boron is limited due to the requirement that the MTC must remain negative over the fuel cycle. Too much boron, typically greater than 1,300 ppm at full power, will make the MTC positive. The optimal design of burnable absorbers is key to the feasibility of this extended cycle and low boron core below the design limit of peak pin power. New concepts for burnable absorbers include changing the materials and geometry in the burnable absorber. k{sub inf}, peaking factor, MTC, and control rod worth of new BAs were compared with those of the conventional BA. A new enriched Er-167 based BA has been proposed and, from three test cases, it was shown that the Erbium burnable absorber is favorable to counterbalance the power peak and Gadolinium burnable absorber is favorable to flattening k{sub inf} trends over burnup.

  20. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y

    2007-06-15

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs.

  1. Strengthened, biaxially textured Ni substrate with small alloying additions for coated conductor applications

    Science.gov (United States)

    Goyal, A.; Feenstra, R.; Paranthaman, M.; Thompson, J. R.; Kang, B. Y.; Cantoni, C.; Lee, D. F.; List, F. A.; Martin, P. M.; Lara-Curzio, E.; Stevens, C.; Kroeger, D. M.; Kowalewski, M.; Specht, E. D.; Aytug, T.; Sathyamurthy, S.; Williams, R. K.; Ericson, R. E.

    2002-11-01

    Fabrication of a biaxially textured, strengthened Ni substrate with small alloying additions of W and Fe is reported. The substrates have significantly improved mechanical properties compared to 99.99% Ni and surface characteristics which are similar to that of 99.99% Ni substrates. High quality oxide buffer layers can be deposited on these substrates without the need for any additional surface modification steps. Grain boundary misorientation distributions obtained from the substrate show a predominant fraction of low-angle grain boundaries. A high critical current density, Jc, of 1.9 MA/cm 2 at 77 K, self-field is demonstrated on this substrate using a multilayer configuration of YBCO/CeO 2/YSZ/Y 2O 3/ Ni-3at.%W-1.7at.%Fe. This translates to a Ic/width of 59 A/cm at 77 K and self-field. Jc at 0.5 T is reduced by only 21% indicating strongly-linked grain boundaries in the YBCO film on this substrate.

  2. Recent improvements of reactor physics codes in MHI

    Energy Technology Data Exchange (ETDEWEB)

    Kosaka, Shinya, E-mail: shinya-kosaka@mhi.co.jp; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki [Mitsubishi Heavy Industries, Ltd. (Japan)

    2015-12-31

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  3. ILK statement about higher-level conclusion drawn from the event at KKP-2 in connection with the 2001 revision; ILK-Stellungnahme zu uebergeordneten Schlussfolgerungen aus den Ereignissen in KKP 2 in Zusammenhang mit der Revision 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-09-01

    The International Committee on Nuclear Technology (ILK), in its most recent statement, deals with general higher-level aspects resulting from the existing documentation of the events at the Philippsburg 2 nuclear power station (KKP-2) in connection with the 2001 revision. Consequences are treated with respect to safty, safety culture, and the observance of criteria. The ILK feels that the events at Philippsburg are reason enough, above and beyond their direct consequences, to deal with a number of fundamental aspects, and also is of the opinion that a number of useful findings have been made which pertain not only to this plant. Among others, these are the following findings: Safety margins are integral parts of the concept of safety staggered in depth. They must be created to the extent necessary, and they must be available. A safety-oriented organization of the operator must be set up in a planned way, and must be expanded continuously. The relationship between the operator and the regulatory authority must be characterized by the objective of achieving constructive, useful solutions, but must also contain clear requirements to be met by the operator. Requirements and specifications must be formulated unequivocally. As far as the specific events at KKP-2 are concerned, the ILK notes that the operator outlined and analyzed the sequence of events in a way easy to follow, and that the measures derived are adequate. (orig.) [German] Die Internationale Laenderkommission Kerntechnik (ILK) behandelt in ihrer aktuellen Stellungnahme generelle, uebergeordnete Gesichtspunkte, die sich auf den vorliegenden Unterlagen zu den Ereignissen im Kernkraftwerk Philippsburg 2 (KKP-2) im Zusammenhang mit der Revision 2001 ergeben. Es werden Folgerungen fuer die Sicherheit, Sicherheitskultur und die Einhaltung von Vorgaben behandelt. Die ILK ist der Auffassung, dass die Vorkommnisse in Philippsburg Anlass sind, ueber die unmittelbaren Konsquenzen hinaus, einige grundlegende

  4. The practical aspects of discontinuation of nucler plant operation; Die Stillegung kerntechnischer Anlagen in der Praxis

    Energy Technology Data Exchange (ETDEWEB)

    Eickelpasch, N. [Kernkraftwerk Gundremmingen Betriebsgesellschaft mbh (Germany)]|[Versuchsatomkraftwerk Kahl (Germany); Guethoff, B. [Preussen Elektra Kernkraft, Kernkraftwerk Wuergassen (Germany); Jungmann, C.R. [Wiederaufbereitungsanlage, Karlsruhe (Germany); Rittscher, D. [Energiewerke Nord GmbH, Lubmin (Germany); Steiner, H. [Kernkraftwerk Gundremmingen Betriebsgesellschaft mbh (Germany)

    1999-05-01

    Dosisbelastung fuer das Personal, einer ueberschaubaren Menge an radioaktiven Abfaellen und unter vernuenftigen wirtschaftlichen Randbedingungen durchgefuehrt werden kann. Die zu treffende Entscheidung zwischen `Sofortigem Abbau` und `Abbau nach Sicherem Einschluss` kann dabei nicht pauschal getroffen werden, sondern haengt von den spezifischen Randbedingungen der Anlage ab - dazu gehoeren die Personalsituation, die Infrastruktur am Ort, Moeglichkeiten zur Endlagerung, aber vor allem finanztechnische Ueberlegungen. Benoetigt werden kuenftig vor allem intelligente Zerlegetechniken - in dem Sinne, dass bekannte Verfahren auch in feindlicher Umgebung, wie hohem Strahlenfeld und grosser Wassertiefe - moeglichst zuverlaessig arbeiten und bei Versagen Reparaturmoeglichkeiten eingeplant werden koennen. Die Behauptung, dass die Stillegung eines Kernkraftwerkes nur mit einem hohen Aufwand an Fernhantiertechnik moeglich sei, hat sich nicht bestaetigt - ihr Einsatz ist hauptsaechlich auf Unterwasserarbeiten beschraenkt. Gerade im Hinblick auf die Endlagerproblematik ist es wichtig, die Menge des radioaktiven Abfalls zu reduzieren. Dafuer muessen dringend leistungsfaehige Dekontaminationsverfahren entwickelt werden, wie z.B. die chemische Dekontamination des Primaerkreises oder das Strahlen von Anlagenteilen mit Stahlkies. (orig.)

  5. On-site Interim Stores for Decommissioning Waste; Standort-Zwischenlager fuer Rueckbauabfaelle

    Energy Technology Data Exchange (ETDEWEB)

    Geiser, H. [Wissenschaftlich - Technische Ingenieurberatung GmbH (WTI), Juelich (Germany)

    2003-05-01

    Periods of interim storage of radioactive waste above ground perhaps up until 2040 must be bridged in case a repository will not be available in time. Ongoing operation of nuclear power plants and, especially, the increasing rate of decommissioning and demolition of power reactors add to the need to plan for the management of waste arising in operation and demolition. Most of the existing interim storage capacity has been earmarked for waste arising in plant operation. It will become necessary to create additional interim storage capacity in order to allow plants to be decommissioned and demolished speedily and, at the same time, make the necessary provisions for interim storage pending final storage. Government institutions and research centers (FZK, FZJ, VKTA, and EWN) created new storage capacity in recent years. The waste arising from decommissioning and demolition of the Hanau nuclear plants will also be emplaced in a new on-site interim store. At a number of sites where power reactors are going to be decommissioned and demolished, operators also are planning for new interim storage capacity for radioactive waste. An overview is given of the on-site interim stores newly built and the new interim stores currently in the planning phase on various sites where plants are to be dismantled. (orig.) [German] Fuer den Fall, dass ein Endlager nicht rechtzeitig zur Verfuegung steht, sind Zwischenlagerzeitraeume fuer die oberirdische Lagerung von radioaktiven Abfaellen u.U. bis 2040 zu ueberbruecken. Durch den laufenden Betrieb der Kernkraftwerke und insbesondere durch den jetzt verstaerkt einsetzenden Rueckbau der Leistungsreaktoren ergibt sich ein zusaetzlicher Planungszwang als Vorsorge fuer die Entsorgung der Betriebs- und Rueckbauabfaelle. Die bestehenden Zwischenlagerkapazitaeten sind weitgehend fuer die Aufnahme von Betriebsabfaellen verplant. Es wird zukuenftig notwendig werden, weitere Zwischenlagerkapazitaeten zu schaffen, damit ein zuegiger Rueckbau moeglich

  6. Optimization potential in maintenance; Optimierungspotenzial in der Instandhaltung

    Energy Technology Data Exchange (ETDEWEB)

    Janisch, H. [EnBW Kraftwerke AG, Kernkraftwerk Philippsburg (Germany)

    2001-04-01

    hohen Zuverlaessigkeit der Anlagen sind diese kapazitaetsbestimmenden Einfluesse auf die Erzeugungskosten aber praktisch ausgeschoepft. Somit steht der Kostenterm im Blickpunkt von Optimierungsansaetzen, d.h. bei den bestehenden Kernkraftwerken insbesondere die Instandhaltungskosten. Bei einem hohen und sich weiter entwickelnden Sicherheitsniveau kann dabei mit den zunehmenden Betriebserfahrungen von der vorbeugenden Instandhaltung auf die kostenoptimierte zustandsorientierte Instandhaltung uebergegangen werden. Weitere Massnahmen fuer fortgeschrittene Instandhaltungsstrategien betreffen die Arbeitsorganisation, das Personalmanagement sowie standortuebergreifende Planungen fuer den Personal- und Anlageneinsatz. Wichtige Komponenten sind EDV-gestuetzte Betriebsfuehrungssysteme, mit denen die komplexen Aufgaben geplant, durchgefuehrt, kontrolliert und analysiert werden koennen. Ziel dieser Massnahmen ist ein weiter optimierter Betrieb der Kernkraftwerke mit hoechster, dem technischen Fortschritt folgender, Sicherheit, bei demonstrierten wettbewerbsfaehigen Stromerzeugungskosten. (orig.)

  7. Masaż Shantala – charakterystyka i spos ób wykonania

    Directory of Open Access Journals (Sweden)

    Iwona Wilk

    2015-10-01

    Full Text Available Masaż pozytywnie wpływa na organizm człowieka, niezależnie od wieku. Wspomaga pracę serca, układu oddechowego i odporność organizmu. W masażu poprzez dotyk stymulujemy receptory czucia powierzchownego zlokalizowane w skórze i dzięki temu możemy zainicjować rozwój motoryki dziecka. Rodzice powinni wykonywać masaż Shantala, ponieważ w prosty sposób mogą wspomóc prawidłowy rozwój swojego dziecka. Prezentowany w artykule rodzaj masażu polega na stosowaniu wyłącznie techniki głaskania powierzchownego na poszczególnych częściach ciała, które masuje się w odpowiedniej kolejności, we właściwym kierunku i w określonym tempie. Ruchy są proste i łatwe do odtworzenia dla rodzica. Masaż Shantala wykonywany systematycznie i prawidłowo pozytywnie oddziałuje na psychikę dziecka. Uspokaja, ułatwia zasypianie, zmniejsza objawy kolki. Przede wszystkim pomaga w wytworzeniu więzi pomiędzy rodzicami a dzieckiem, dając poczucie bezpieczeństwa i wsparcia. Cechą charakterystyczną tego rodzaju masażu jest fakt, iż odbiorca, czyli dziecko, i wykonawca, czyli rodzic, czerpią pozytywne doznania płynące z dotyku i bliskości. Masaż umożliwia obu stronom wyciszenie, uspokojenie i chwilę relaksu.

  8. Simplified containment event tree analysis for the Sequoyah Ice Condenser containment

    Energy Technology Data Exchange (ETDEWEB)

    Galyean, W.J.; Schroeder, J.A.; Pafford, D.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-12-01

    An evaluation of a Pressurized Water Reactor (PWR) ice condenser containment was performed. In this evaluation, simplified containment event trees (SCETs) were developed that utilized the vast storehouse of information generated by the NRC's Draft NUREG-1150 effort. Specifically, the computer programs and data files produced by the NUREG-1150 analysis of Sequoyah were used to electronically generate SCETs, as opposed to the NUREG-1150 accident progression event trees (APETs). This simplification was performed to allow graphic depiction of the SCETs in typical event tree format, which facilitates their understanding and use. SCETs were developed for five of the seven plant damage state groups (PDSGs) identified by the NUREG-1150 analyses, which includes: both short- and long-term station blackout sequences (SBOs), transients, loss-of-coolant accidents (LOCAs), and anticipated transient without scram (ATWS). Steam generator tube rupture (SGTR) and event-V PDSGs were not analyzed because of their containment bypass nature. After being benchmarked with the APETs, in terms of containment failure mode and risk, the SCETs were used to evaluate a number of potential containment modifications. The modifications were examined for their potential to mitigate or prevent containment failure from hydrogen burns or direct impingement on the containment by the core, (both factors identified as significant contributors to risk in the NUREG-1150 Sequoyah analysis). However, because of the relatively low baseline risk postulated for Sequoyah (i.e., 12 person-rems per reactor year), none of the potential modifications appear to be cost effective. 15 refs., 10 figs. , 17 tabs.

  9. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko, M.; Antti, D.; Hanna, R.; Timo, V. [VTT Processes, (Finland)

    2004-07-01

    TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable part in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but the rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. (authors)

  10. The significance of biometric parameters in determining anterior teeth width

    Directory of Open Access Journals (Sweden)

    Strajnić Ljiljana

    2013-01-01

    Full Text Available Background/Aim. An important element of prosthetic treatment of edentulous patients is selecting the size of anterior artificial teeth that will restore the natural harmony of one’s dentolabial structure as well as the whole face. The main objective of this study was to determine the correlation between the inner canthal distance (ICD and interalar width (IAW on one side and the width of both central incisors (CIW, the width of central and lateral incisors (CLIW, the width of anterior teeth (ATW, the width between the canine cusps (CCW, which may be useful in clinical practice. Methods. A total of 89 subjects comprising 23 male and 66 female were studied. Their age ranged from 19 to 34 years with the mean of 25 years. Only the subjects with the preserved natural dentition were included in the sample. All facial and intraoral tooth measurements were made with a Boley Gauge (Buffalo Dental Manufacturing Co., Brooklyn NY, USA having a resolution of 0.1mm. Results. A moderate correlation was established between the interalar width and combined width of anterior teeth and canine cusp width (r = 0.439, r = 0.374. A low correlation was established between the inner canthal distance and the width of anterior teeth and canine cusp width (r = 0.335, r = 0.303. The differences between the two genders were highly significant for all the parameters (p < 0.01. The measured facial distances and width of anterior teeth were higher in men than in women. Conclusion. The results of this study suggest that the examined interalar width and inner canthal distance cannot be considered reliable guidelines in the selection of artificial upper anterior teeth. However, they may be used as a useful additional factor combined with other methods for objective tooth selection. The final decision should be made while working on dentures fitting models with the patient’s consent.

  11. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station; Analisis de eventos internos para la Unidad 1 de la Central Nucleolelectrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1993-07-01

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  12. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results; Analisis de eventos internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Cuantificacion de secuencias de accidente y resultados

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1994-07-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the {sup I}nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

  13. Aspiration tests in aqueous foam using a breathing simulator

    Energy Technology Data Exchange (ETDEWEB)

    Archuleta, M.M.

    1995-12-01

    Non-toxic aqueous foams are being developed by Sandia National Laboratories (SNL) for the National Institute of Justice (NIJ) for use in crowd control, cell extractions, and group disturbances in the criminal justice prison systems. The potential for aspiration of aqueous foam during its use and the resulting adverse effects associated with complete immersion in aqueous foam is of major concern to the NIJ when examining the effectiveness and safety of using this technology as a Less-Than-Lethal weapon. This preliminary study was designed to evaluate the maximum quantity of foam that might be aspirated by an individual following total immersion in an SNL-developed aqueous foam. A.T.W. Reed Breathing simulator equipped with a 622 Silverman cam was used to simulate the aspiration of an ammonium laureth sulfate aqueous foam developed by SNL and generated at expansion ratios in the range of 500:1 to 1000:1. Although the natural instinct of an individual immersed in foam is to cover their nose and mouth with a hand or cloth, thus breaking the bubbles and decreasing the potential for aspiration, this study was performed to examine a worst case scenario where mouth breathing only was examined, and no attempt was made to block foam entry into the breathing port. Two breathing rates were examined: one that simulated a sedentary individual with a mean breathing rate of 6.27 breaths/minute, and one that simulated an agitated or heavily breathing individual with a mean breathing rate of 23.7 breaths/minute. The results of this study indicate that, if breathing in aqueous foam without movement, an air pocket forms around the nose and mouth within one minute of immersion.

  14. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 2: Accident and Thermal Fluids Analysis PIRTs

    Energy Technology Data Exchange (ETDEWEB)

    Ball, Sydney J [ORNL; Corradini, M. [University of Wisconsin; Fisher, Stephen Eugene [ORNL; Gauntt, R. [Sandia National Laboratories (SNL); Geffraye, G. [CEA, France; Gehin, Jess C [ORNL; Hassan, Y. [Texas A& M University; Moses, David Lewis [ORNL; Renier, John-Paul [ORNL; Schultz, R. [Idaho National Laboratory (INL); Wei, T. [Argonne National Laboratory (ANL)

    2008-03-01

    An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: (1) normal operation (including some reactor physics aspects), (2) general loss of forced circulation (G-LOFC), (3) pressurized loss-of-forced circulation (P-LOFC), (4) depressurized loss-of-forced circulation (D-LOFC), (5) air ingress (following D-LOFC), (6) reactivity transients - including anticipated transients without scram (ATWS), (7) processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and (8) steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: (1) core coolant bypass flows (normal operation), (2) power/flux profiles (normal operation), (3) outlet plenum flows (normal operation), (4) reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation and accidents), (5) fission product release related to the transport of silver (normal operation), (6)emissivity aspects for the vessel and reactor cavity cooling system (G-LOFC), (7) reactor vessel cavity air circulation and heat transfer (G-LOFC), and (8)convection/radiation heating of upper vessel area (P-LOFC).

  15. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Papukchiev, Angel [GRS mbH Forschungsinstitute, Garching (Germany); Schaefer, Anselm [ISaR GmbH, Garching (Germany)

    2008-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  16. Causes, consequences, and therapy of the Radiophobia syndrome; Ursachen, Folgen und Therapie des Radiophobie-Syndroms

    Energy Technology Data Exchange (ETDEWEB)

    Becker, K.

    2004-03-01

    The final storage of high-level radioactive waste, which is said to be still open while, in fact, it was solved technically a long time ago and is only being blocked for political reasons, as well as alleged technical risks of German nuclear power plants which have never been demonstrated or proven, are listed again and again as grounds for opting out of the use of nuclear power. There is hardly any doubt that one of the main causes underlying also these arguments, and thus the main reason for the insufficient public acceptance of nuclear power in Germany at the present time as a safe, inexpensive, and non-polluting source of primary energy, is the widespread fear of radiation (radiophobia). Consequently, solutions proposed for successfully managing this radiophobia must be examined. Continued scientific studies of the subject do not seem to be promising, as funds are available at present only for continuing the search for negative biological effects. Important preconditions for a change in attitude are the appropriate initiatives to be taken by the relatively small number of sufficiently independent experts of proven scientific repute. Initiatives of this kind can now be observed in numerous countries and regions in the world. It must be pointed out in this connection, as is underlined again and again by experienced experts, that risk acceptance is not a matter of factual arguments, but of emotions. Psychological and pedagogic sensitivity certainly are important elements in changing public opinion in the interest of a more realistic assessment of the radiation risk and the acceptance of nuclear power. (orig.) [German] Die angeblich noch offene, tatsaechlich aber laengst technisch geloeste und nur politisch blockierte Frage der Endlagerung hochradioaktiver Abfaelle, ebenso wie vorgebliche, tatsaechlich aber nie nachgewiesene technische Risiken der deutschen Kernkraftwerke werden immer wieder als Ausstiegsgruende fuer die Kernenergie genannt. Es bestehen kaum

  17. Reports within the area of nuclear power plant instrumentation: Part 1: Laboratory test of analogue and digital instrument components. Part 2: Dynamic deviations in reactor pressure water level signals caused by sensing lines

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, Bengt-Goeran [GSE Power Systems AB, Nykoeping (Sweden)

    2004-11-01

    type TDE220. The transmitters exhibited deviating dynamics during ordinary sensor tests. The laboratory test confirms the observed deviation in comparison with transmitters of other types. The construction with Bourdon tube is judged to be the reason for the deviations. The report also presents results from trouble shooting with steam pressure transmitters at KKM (Kernkraftwerk Muehleberg m Switzerland). It was possible to identify the intermittent sensor error with the aid of controlled pressure changes. Service of the transmitter pointed out a crack on the electronic filter unit. This was judged to be the reason for the intermittent signal interrupts. Finally, two possibilities used at KKM to investigate the dynamics of temperature sensors are described. Both methods are based on artificial cooling of the sensor. One of them is applied during power operation of the plant and the other during outage. (The present report is a translation of the Swedish language report SKI-R--03-07, published Dec 2002)

  18. VAK Kahl - decommissioning and demolition continued under new auspices; VAK Kahl - Fortsetzung des Rueckbaus unter neuem Vorzeichen

    Energy Technology Data Exchange (ETDEWEB)

    Hackel, W.; Runge, H. [RWE NUKEM GmbH, Alzenau (Germany)

    2001-11-01

    The Kahl experimental nuclear power station (VAK), the first German nuclear power plant, was decommissioned after 25 years of operation (1961 to 1985). The BWR plant generated approx. 2 million kWh of electricity in 150,000 hours of operation at a gross power of MWe. After the operator, VAK GmbH, had filed an application for decommissioning, the first of four decommissioning permits was issued in 1988. The plant is to be demolished completely so that the site will no longer be within the scope of the Atomic Energy Act. By 2001, demolition work covered by the first decommissioning permit had been finished, also the 2nd and 3rd decommissioning permits had largely been completed, and work under the 4th decommissioning permit had been begun. To acquire technical and organization experience and know-how, the decommissioning and demolition phases are accompanied by research and development work carried out by the operators and by VAK shareholders RWE and E.ON. After the bulk of the work had been completed, the radioactive inventory had been removed from the plant, and the end of the project was in sight, RWE NUKEM GmbH was commissioned to carry on. The main objectives now are speedy completion of the jobs still to be finished, further development for other projects of the know-how acquired, and job protection. The main work still to be carried out includes dismantling of systems no longer needed and of the biological shield as well as decontamination of building structures accompanied by the clearance of buildings and open areas for subsequent conventional demolition. The waste arising will be packaged in accordance with its classification, and will be removed into interim storage or managed in the conventional way. The project is to be completed in the 3rd quarter of 2006. (orig.) [German] Das Versuchsatomkraftwerk Kahl (VAK) mit Siedewasserreaktor, das erste deutsche Kernkraftwerk, wurde nach 25 Betriebsjahren (1961 bis 1985) stillgelegt. Die Siedewasserreaktoranlage

  19. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  20. Implementing standards for the interoperability among healthcare providers in the public regionalized Healthcare Information System of the Lombardy Region.

    Science.gov (United States)

    Barbarito, Fulvio; Pinciroli, Francesco; Mason, John; Marceglia, Sara; Mazzola, Luca; Bonacina, Stefano

    2012-08-01

    Information technologies (ITs) have now entered the everyday workflow in a variety of healthcare providers with a certain degree of independence. This independence may be the cause of difficulty in interoperability between information systems and it can be overcome through the implementation and adoption of standards. Here we present the case of the Lombardy Region, in Italy, that has been able, in the last 10 years, to set up the Regional Social and Healthcare Information System, connecting all the healthcare providers within the region, and providing full access to clinical and health-related documents independently from the healthcare organization that generated the document itself. This goal, in a region with almost 10 millions citizens, was achieved through a twofold approach: first, the political and operative push towards the adoption of the Health Level 7 (HL7) standard within single hospitals and, second, providing a technological infrastructure for data sharing based on interoperability specifications recognized at the regional level for messages transmitted from healthcare providers to the central domain. The adoption of such regional interoperability specifications enabled the communication among heterogeneous systems placed in different hospitals in Lombardy. Integrating the Healthcare Enterprise (IHE) integration profiles which refer to HL7 standards are adopted within hospitals for message exchange and for the definition of integration scenarios. The IHE patient administration management (PAM) profile with its different workflows is adopted for patient management, whereas the Scheduled Workflow (SWF), the Laboratory Testing Workflow (LTW), and the Ambulatory Testing Workflow (ATW) are adopted for order management. At present, the system manages 4,700,000 pharmacological e-prescriptions, and 1,700,000 e-prescriptions for laboratory exams per month. It produces, monthly, 490,000 laboratory medical reports, 180,000 radiology medical reports, 180

  1. The basic research on the CDA initiation phase for a metallic fuel FBR

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Go; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan); Kawada, Ken-ichi; Niwa, Hazime

    1998-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

  2. Barrier efficiency of sponge-like La{sub 2}Zr{sub 2}O{sub 7} buffer layers for YBCO-coated conductors

    Energy Technology Data Exchange (ETDEWEB)

    Molina, Leopoldo; Tan, Haiyan; Biermans, Ellen; Verbeeck, Jo; Bals, Sara; Tendeloo, Gustaaf Van [EMAT, University of Antwerp, Groenenborgerlaan 171, BE-2020 Antwerp (Belgium); Batenburg, Kees J, E-mail: leopoldo.molina-luna@ua.ac.be [Vision Lab, University of Antwerp, Universiteitsplein 1, BE-2020 Wilrijk (Belgium)

    2011-06-15

    Solution derived La{sub 2}Zr{sub 2}O{sub 7} films have drawn much attention for potential applications as thermal barriers or low-cost buffer layers for coated conductor technology. Annealing and coating parameters strongly affect the microstructure of La{sub 2}Zr{sub 2}O{sub 7}, but different film processing methods can yield similar microstructural features such as nanovoids and nanometer-sized La{sub 2}Zr{sub 2}O{sub 7} grains. Nanoporosity is a typical feature found in such films and the implications for the functionality of the films are investigated by a combination of scanning transmission electron microscopy (STEM), electron energy-loss spectroscopy (EELS) and quantitative electron tomography. Chemical solution based La{sub 2}Zr{sub 2}O{sub 7} films deposited on flexible Ni-5 at.%W substrates with a {l_brace}100{r_brace}(001) biaxial texture were prepared for an in-depth characterization. A sponge-like structure composed of nanometer-sized voids is revealed by high-angle annular dark-field scanning transmission electron microscopy in combination with electron tomography. A three-dimensional quantification of nanovoids in the La{sub 2}Zr{sub 2}O{sub 7} film is obtained on a local scale. Mostly non-interconnected highly faceted nanovoids compromise more than one-fifth of the investigated sample volume. The diffusion barrier efficiency of a 170 nm thick La{sub 2}Zr{sub 2}O{sub 7} film is investigated by STEM-EELS, yielding a 1.8 {+-} 0.2 nm oxide layer beyond which no significant nickel diffusion can be detected and intermixing is observed. This is of particular significance for the functionality of YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} coated conductor architectures based on solution derived La{sub 2}Zr{sub 2}O{sub 7} films as diffusion barriers.

  3. Fuel rod under power oscillations; calculations with the ENIGMA code

    Energy Technology Data Exchange (ETDEWEB)

    Ranta-Puska, Kari

    1999-05-15

    Power oscillations in a BWR may result from a series of events starting from a re-circulation pump trip or can be initiated during start-up at low-flow conditions by other perturbations. Whole core and regional oscillations have been observed. Severe consequences may be anticipated if the instability diverges and the reactor protection system fails (no scram) in all phases of the incident (ATWS). Power peaks higher than ten times of the pre-transient power level have been speculated to appear. Low-magnitude oscillations have been observed at the TVO plant, Olkiluoto 1987, and at the Lasalle-2 plant, 1988, and in other BWRs world-wide. Typically, a boiling water reactor has an unstable operational point at low flow and high power conditions. The physical phenomenon behind the instability is density wave oscillations leading to boiling boundary oscillations and void fraction fluctuations across the heated channel. These in turn, make the fission power vary. The typical frequency of the oscillations seems to be of the order of 0.5 Hz, and thus the power peak for a fuel rod is considerably wider than a RIA-pulse, for instance. Large oscillations can result in elevated fuel temperatures, accelerated fission gas release and additional internal loads on the cladding. These effects may be more severe for a high burnup rod with a large fission gas inventory and a closed gap. Therefore, an experiment has been proposed to be conducted at Halden reactor for simulating the fuel rod response under power oscillations. As there is lack of knowledge also on the relevant boundary conditions, pre-calculations with various input options have been performed and are further suggested. Calculations with FRAPTRAN code have shown the importance of the cladding-coolant heat transfer to the fuel temperature. The applicability of the ENIGMA code to this kind of transients was confirmed. To support the planning of the proposed Halden test, estimates on fuel and cladding temperatures as well as

  4. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  5. Design of an Actinide-Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low-Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Weaver, Kevan Dean; Davis, Cliff Bybee; MIT folks

    2000-07-01

    The purpose of this Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) University Research Consortium (URC) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, material compatibility, plant engineering, and coolant activation. In the area of core neutronic design, the reactivity vs. burnup and discharge isotopics of both non-fertile and fertile fuels were evaluated. An innovative core for pure actinide burning that uses streaming, fertile-free fuel assemblies was studied in depth. This particular core exhibits excellent reactivity performance upon coolant voiding, even for voids that occur in the core center, and has a transuranic (TRU) destruction rate that is comparable to the proposed accelerator transmutation of waste (ATW) facility. These studies suggest that a core can be designed to achieve a long life while maintaining safety and minimizing waste. In the area of material compatibility studies, an experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials has been designed and built at the INEEL. The INEEL forced-convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The corrosion cell is being used to test steel that is commercially available in the United States to temperatures above 650°C. Progress in plant engineering was made for two reactor concepts, one utilizing an indirect cycle with heat exchangers and the other utilizing a direct-contact steam cycle. The evaluation of the

  6. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko Miettinen; Timo Vanttola; Hanna Raety; Antti Daavittila [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

    2005-07-01

    Full text of publication follows: TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four-equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one-dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five-equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. In general questions could be raised, what advantages are seen with the new internal coupling in comparison with the earlier realised parallel coupling, and which advantages may be seen in building the realtor physical model on the basis of the old code, developed since 1970's. The internal coupling allows

  7. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    studied to assess the possibilities for using three-dimensional cores in training simulators. The core model results have been compared with the Loviisa WWER-type plant measurement data in steady state and in some transients. Hypothetical control rod withdrawal, ejection and boron dilution transients have been calculated with various three-dimensional core models for the Loviisa WWER-440 core. Several ATWS analyses for the WWER-1000/91 plant have been performed using the three-dimensional core model. In this context, the results of APROS have been compared in detail with the results of the HEXTRAN code. The three-dimensional Olkiluoto BWR-type core model has been used for transient calculation and for severe accident re-criticality studies. The one-dimensional core model is at present used in several plant analyser and training simulator applications and it has been used extensively for safety analyses in the Loviisa WWER-440 plant modernisation project. (orig.) 75 refs. The thesis includes also eight previous publications by author

  8. Expert report of ENSI on the request of EKKM AG for a general license - Project 'New nuclear power plant to replace the Muehleberg plant'; Gutachten des ENSI zum Rahmenbewilligungsgesuch der EKKM AG. Neubauprojekt Ersatzkernkraftwerk Muehleberg

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-09-15

    The 'Ersatz Kernkraftwerk Muehleberg AG' (EKKM) Company submitted to the Swiss Federal Inspectorate of Nuclear Safety (ENSI) a request for a general license for a new power plant to be built near to the Muehleberg power plant. According to the law, all damage risks with a probability higher than 10{sup -4}/a must be taken into account through protection measures. The considered risks concern the power plant itself as well as the population in the neighbourhood and the environment. The purpose of the general license is to demonstrate that the site chosen for the foreseen power plant is acceptable and that the risks can be counteracted through adequate measures. The buildings of the power plant and their partition on the left side of the Aare River are briefly described. The reactor is a Light Water Reactor of third generation with a maximum electrical power of 1.6 GW{sub el}. The European Pressurized water Reactor, which is of the same power class, can be taken as an example for planning that will follow. The main cooling is provided by one or, if needed, two cooling towers using a hybrid system of water evaporation and air heating, what reduces the plume at the exit of the cooling towers. The population density in the neighbourhood of the power plant is low; it is demonstrated that, in the case of a very unlikely severe accident in the power plant, the people in the neighbourhood can be evacuated quickly. Then, numerous types of possible accidents in the neighbourhood of the power plant are analyzed in order to settle their possible negative influence on the operation of the power plant: bursting of gas containers on the neighbouring roads and railways, fires of all types of hydrocarbons, air pollution through chloride gas, etc. The check by ENSI of the EKKM studies on the potential danger for the power plant through neighbouring industrial plants, roads or railways demonstrated that none of the considered accidents presents an unacceptable risk for the

  9. Expert report of ENSI on the request of EKKB AG for a general license - Project 'New nuclear power plant to replace the Beznau plant'; Gutachten des ENSI zum Rahmenbewilligungsgesuch der EKKB AG. Neubauprojekt Ersatzkernkraftwerk Beznau

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-09-15

    The 'Ersatz Kernkraftwerk Beznau AG' (EKKB) Company submitted to the Swiss Federal Inspectorate of Nuclear Safety (ENSI) a request for a general license for a new power plant to be built near to the Beznau power plants. According to the law, all damage risks with a probability higher than 10{sup -4}/a must be taken into account through protection measures. The considered risks concern the power plant itself as well as the population in the neighbourhood and the environment. The purpose of the general license is to demonstrate that the site chosen for the foreseen power plant is acceptable and that the risks can be counteracted through adequate measures. The buildings of the power plant and their partition on the Beznau Island in the Aare River are briefly described. The reactor is a Light Water Reactor of third generation with a maximum electrical power of 1450 MW{sub el} {+-}20%. The main cooling is provided by a hybrid system of water evaporation and air heating, what reduces the plume at the exit of the cooling tower. First, it is demonstrated that, in the case of a very unlikely severe accident in the power plant, the people in the neighbourhood can be evacuated quickly. Then, numerous types of possible accidents in the neighbourhood of the power plant are analyzed in order to settle their possible negative influence on the operation of the power plant: bursting of gas containers on the neighbouring roads and railways, fires of all types of hydrocarbons, air pollution through chloride gas, etc. The check by ENSI of the EKKB studies on the potential danger for the power plant through neighbouring industrial plants, roads or railways demonstrated that none of the considered accidents presents an unacceptable risk for the power plant: on the one hand, these plants are located too far from the power plant, so that a sensible injury to the power plant safety can be excluded; on the other, the protection of the power plant can be guaranteed through

  10. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    approach to the PB-FHR annular pebble bed core cooled by fluoride salt mixtures generated a model that is called Pod. Pod. was used to show the resilience of the PB-FHR core to generation of hot spots or cold spots, due to the effect of buoyancy on the flow and temperature distribution in the packed bed. Pod. was used to investigate the PB-FHR response to ATWS transients. Based on the functional requirements for the core, Pod. was used to generate an optimized design of the flow distribution in the core. An analysis of natural circulation loops cooled by single-phase Boussinesq fluids is presented here, in the context of reactor design that relies on natural circulation decay heat removal, and design of scaled experiments. The scaling arguments are established for a transient natural circulation loop, for loops that have long fluid residence time, and negligible contribution of fluid inertia to the momentum equation. The design of integral effects tests for the loss of forced circulation (LOFC) for PB-FHR is discussed. The special case of natural circulation decay heat removal from a pebble bed reactor was analyzed. A way to define the Reynolds number in a multi-dimensional pebble bed was identified. The scaling methodology for replicating pebble bed friction losses using an electrically resistance heated annular pipe and a needle valve was developed. The thermophysical properties of liquid fluoride salts lead to design of systems with low flow velocities, and hence long fluid residence times. A comparison among liquid coolants for the performance of steady state natural circulation heat removal from a pebble bed was performed. Transient natural circulation experimental data with simulant fluids for fluoride salts is given here. The low flow velocity and the relatively high viscosity of the fluoride salts lead to low Reynolds number flows, and a low Reynolds number in conjunction with a sufficiently high coefficient of thermal expansion makes the system susceptible to