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Sample records for atw schnellstatistik kernkraftwerke

  1. Analysis of SBO ATWS for Maanshan PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Che-Hao; Chen, Shao-Wen [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Nuclear and New Energy Education and Research Foundation, Hsinchu, Taiwan (China); Lin, Hao-Tzu [Atomic Energy Council, Taoyuan, Taiwan (China). Inst. of Nuclear Energy Research

    2015-11-15

    Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

  2. 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.

  3. ATW system impact on high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Arthur, E.D.

    1992-12-01

    This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products.

  4. 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.

  5. A Los Alamos concept for accelerator transmutation of waste and energy production (ATW)

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-31

    This document contains the diagrams presented at the ATW (Accelerator Transmutation of Waste and Energy Production) External Review, December 10-12, 1990, held at Los Alamos National Laboratory. Included are the charge to the committee and the presentations for the committee`s review. Topics of the presentations included an overview of the concept, LINAC technology, near-term application -- high-level defense wastes (intense thermal neutron source, chemistry and materials), advanced application of the ATW concept -- fission energy without a high-level waste stream (overview, advanced technology, and advanced chemistry), and a summary of the research issues.

  6. Study of safety relief valve operation under ATWS conditions. [Supercritical flow

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Whitten, S.D.

    1979-09-01

    In March 1979, the NRC published a report (NUREG/CR-0687) prepared by the Energy Technology Engineering Center (ETEC-TDR-78-19). That report presented a literature survey which updated earlier NRC studies of saturated or subcooled water flow through relief valves, under ATWS conditions. This supplement expands upon that search to include supercritical steam-water flow. No applicable data for the supercritical conditions were found, nor were any newer data on saturated or subcooled conditions uncovered. This supplement also updates a look for facilities currently capable of simultaneously imposing all ATWS conditions upon test relief valves. Results confirmed the negative findings of NUREG/CR-0687.

  7. Study of safety relief valve operation under ATWS conditions. [Super critical flow

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Whitten, S.D.

    1979-07-25

    In March 1979, ETEC published as ETEC-TDR-78-19 a search which updated earlier NRC studies of saturated or subcooled water flow through relief valves, under ATWS conditions. This Supplement expands upon that search to include supercritical steam-water flow. No applicable data for the supercritical conditions was found, nor were any newer data on saturated or subcooled conditions uncovered. The Supplement also updated a look for facilities currently capable of simultaneously imposing all ATWS conditions upon test relief valves. Results confirmed the negative findings of ETEC-TDR-78.19.

  8. 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.

  9. Licensing issues in the context of terrorist attacks on nuclear power plants; Genehmigungsrechtliche Fragen terroristischer Angriffe auf Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Danwitz, T. von

    2002-07-01

    The terrorist attack on the World Trade Center in September 2001 has prompted enhanced nuclear risk awareness among the German population. But in the current public debate about the safety of nuclear power plants in Germany in times of new dimensions of danger, aspects such as the role of the constitutional law, the German Atomic Energy Act, and the regulatory system governing nuclear power plant licensing in the context of protection and safety have not been addressed. The author therefore discusses the German nuclear power plant licensing law and administrative regime, elaborating on the significance attributed in those bodies of law to risks like terrorist attacks on nuclear power plants. (orig./CB) [German] Das allgemeine Risiko von terroristischen Anschlaegen auf Kernkraftwerke ist durch die Ereignisse vom 11. September 2001 wieder verstaerkt in das Bewusstsein der Oeffentlichkeit getreten. Die verfassungsrechtlichen Grundlagen und die atomgesetzliche Einordnung der Risiken von terroristischen Angriffen auf kerntechnische Anlagen bleiben jedoch in der aktuellen Diskussion weithin ungeklaert. Der Beitrag unternimmt es daher, die verfassungs- und verwaltungsrechtliche Bedeutung der Risiken terroristischer Angriffe auf Kernkraftwerke in atomrechtlichen Genehmigungsverfahren zu untersuchen. (orig./CB)

  10. Texture optimization of Ni-5at.%W for coated conductor applications

    Energy Technology Data Exchange (ETDEWEB)

    Witte, Marco

    2014-07-01

    For the application of Ni-5at.%W as substrate for high-temperature superconductors the development of a very intense Cube texture is essential. Therefore, the texture development in Ni-5at.%W was studied in detail by experiments and computer simulations. Cold rolling of the material resulted in a pronounced β-fiber texture which transformed by recrystallization into a strong Cube texture and then sharpened further during subsequent grain growth. The cold rolling process was found to be limited by the occurrence of deformation inhomogeneities to strains of ε = 3-4 and led to the formation of Cube deformation bands. These bands formed very effective nucleation sites during recrystallization if ε > 2.5. The deformation texture evolution could be successfully simulated with the grain interaction model (GIA) when a certain amount of random shear deformation was allowed. During annealing of cold rolled Ni-5at.%W recrystallization nuclei emerged from the deformed Cube bands earlier and with higher frequency than for any other orientation and the Cube volume fraction increased from 0.1% to 50%. Furthermore, the Cube nuclei had the highest mean misorientation of all orientations and thus they had the highest fraction of mobile high angle boundaries (HAGB). Together with the nucleation advantage this led to a bimodal size distribution of large Cube grains and smaller grains of other orientations after recrystallization was completed. If the annealing temperature for recrystallization was too high this nucleation and growth advantage of the Cube grains was reduced. Thus, a two-step annealing process with a lower temperature for recrystallization and a higher temperature for the grain growth stage was found to be beneficial. With the statistical recrystallization texture model (StaRT) the development of the recrystallization texture was simulated based on results of a previous GIA simulation and an experimentally determined nucleation spectrum. A delay function was

  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. 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.

  13. KLE annual report 1987. Kernkraftwerke Lippe-Ems GmbH (KLE). Bericht ueber das Geschaeftsjahr 1987

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    The KLE utility's project since 1982 is the Emsland reactor station. Construction and assembly work could be almost completed in 1987. Pressure testing of the primary loop was done in March; in May the third partial licence was issued, so that commissioning work could be started: The thermal test phase 1 began on September 14, and on October 3 the turbo-generator set was first switched to the 380 kV grid of VEW. The fuel elements for the first reactor core were supplied and after temporary storage in the fuel element pond, loading of the core was done in February/March 1988. The utility's natural uranium stocks amounted to about 509 t after the natural uranium for fuel element fabrication of the first core had been supplied. The continuing downward trend of market prices for nuclear fuels required depreciations and reserves to be made in the nuclear fuel sector acounting to DM 104 millions. Interest expense for outside financing of the Emsland reactor project, amounting to DM 188 millions, have had a major influence on the financial situation in the reporting year. An annual loss of DM 354 millions had to be compensated, which was done within the framework of existing agreements with the Kraftwerksverwaltungs-oHG, Vereinigte Elektrizitaetswerke Westfalen AG, and ELEKTROMARK Kommunales Elektrizitaetswerk Mark AG, Vereinigung der Gesellschafter der Kernkraftwerke Lippe-Ems. The operating personnel is now almost complete and covers 209 persons.

  14. Preliminary Safety Analysis of Anticipated Transient without Scram (ATWS) Events for the Prototype GEN-IV SFR (PGSFR) using MARS-LMR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A safety analysis of ATWS for the recently designed Prototype GEN-IV Sodium Cooled Fast Reactor (PGSFR) was conducted. Unprotected Transient Over-Power (UTOP), Unprotected Loss OF Flow (ULOF), and Unprotected Loss Of Heat Sink (ULOHS) were selected as representative events for the ATWS. In an unprotected condition, the power in the core is only controlled by reactivity feedbacks, which are interacted with the thermal-hydraulic characteristics of the components in the plant. Heat is removed by the steam generator (SG) and decay heat removal system (DHRS). Therefore, the major objectives of the safety analysis of the ATWS events are to investigate the thermal hydraulic characteristics of the DHRS and the SG, the neutron kinetic characteristics of the reactivity feedback, and the interaction between the neutron kinetics, and the thermal-hydraulics during the events.

  15. Spectroscopic classification of Gaia16atw and Gaia16aui with the SEDM (Spectra Energy Distribution Machine) on Palomar 60-inch (P60) telescope

    Science.gov (United States)

    Blagorodnova, N.; Neill, D.; Walters, R.

    2016-07-01

    The Caltech Time Domain Astronomy group reports the classification of Gaia16atw and Gaia16aui, discovered by the Gaia ESA survey. The observations were performed with the Spectral Energy Distribution Machine (SEDM)(http://www.astro.caltech.edu/sedm/, range 350-950nm, spectral resolution R~100) on Palomar 60-inch (P60) telescope.

  16. One year of operational experience with the upgraded and modernized Borssele nuclear power plant; Ein Jahr Erfahrung mit dem nachgeruesteten und modernisierten Kernkraftwerk Borssele

    Energy Technology Data Exchange (ETDEWEB)

    Bongers, J.W.M. [EPZ Kernenergie, Borssele (Netherlands); Wiersema, H.T. [KEMA Nuclear, Arnheim (Netherlands)

    1999-10-01

    The operating experience with the modernized Borssele NPP is excellent. The post upgrade operations showed some minor incidents related to the modernization. Part of the incidents were handled immediately, the others were solved during the 1998 outage. This outage was very well organized and executed, resulting in the shortest outage time ever achieved in Borssele. The plant availability in the first operating period after modernization ranged at over 90%. (orig.) [Deutsch] Mit dem modernisierten Kernkraftwerk Borssele wurden sehr gute Erfahrungen gemacht. Waehrend des Betriebes nach Abschluss der Modernisierungsarbeiten kam es zu einigen geringfuegigen Ereignissen, die durch die Modernisierungsmassnahmen verursacht worden waren. Einige dieser Probleme konnten sofort behoben werden, andere wurden waehrend der Revision im Jahre 1998 geloest. Diese Revision wurde hervorragend organisiert und ausgefuehrt und fuehrte dadurch zu den kuerzesten Ausfallzeiten, die jemals in Borssele erzielt wurden. Die Anlagenverfuegbarkeit lag in der ersten Betriebszeit nach den Umbaumassnahmen bei ueber 90%. (orig.)

  17. Influence of intermediate annealing on the microstructure and texture of Ni-9.3at%W substrates

    Institute of Scientific and Technical Information of China (English)

    Jia-nan Liu; Wei Liu; Guo-yi Tang; Ru-fei Zhu

    2014-01-01

    The effects of intermediate annealing (IA) on the microstructure and texture of Ni-9.3at%W substrates have been investigated by using electron backscattering diffraction and X-ray diffraction. Results suggest that IA can optimize the homogeneity of deformation micro-structure. Higher IA temperatures (without undergoing recrystallization during IA) will increase the copper-type components of deformation texture and improve the content of cube texture after recrystallization. Sharp cube texture (97.2%) can be obtained at the optimum IA tem-perature of 650°C. The mechanism underlying the transition of deformation texture can be interpreted as that IA increases the dislocation slipping ability and suppresses the twinning deformation of Copper orientation in the subsequent rolling process. The observed strengthening of cube texture as a result of IA treatment is presumably attributed to the reduction of noncube nucleation and the optimization of preferential growth surrounding the cube nuclei.

  18. Effect of Initial Surface Quality on Final Roughness and Texture of Annealed Ni-5at.%W Tapes Coated with a Gd2Zr2O7 Buffer Layer

    DEFF Research Database (Denmark)

    Wulff, Anders Christian; Yue, Zhao; Mishin, Oleg;

    2012-01-01

    Surface roughness of Ni-5at.%W tapes in coldrolled and annealed conditions after subsequent deposition of a Gd2Zr2O7 buffer layer has been studied as a function of the polishing grade, taking grain boundary grooving into account. It is found that annealing decreases the initial mean surface rough...

  19. 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

    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....... Fully covered films with a broad thickness range exhibit a high degree of biaxial orientation, similar surface morphology with crack free and nano-size grains microstructure, seemingly independent of neither heat treatment nor thickness. Particularly, we compared the porosity of the film surface...... and body according to surface or cross-sectional observation and Rutherford Backscattering Spectrometry analysis, pointing to inhomogeneous structure through film thickness, i.e., dense in the surface layer but porous in the body. This is attributed to trapped gas generated during either decomposition...

  20. 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).

  1. Test of different measures for the prevention of scaling in the cooling system of Grohnde nuclear power plant; Test verschiedener Massnahmen zur Verhuetung von Kalkablagerungen im Kuehlsystem des Kernkraftwerks Grohnde

    Energy Technology Data Exchange (ETDEWEB)

    Czolkoss, W. [Taprogge GmbH, Wetter (Germany); Jacobi, G.; Schueler, A. [Gemeinschaftskernkraftwerk Grohnde GmbH, Emmerthal (Germany); Fichte, W. [Allianz-Zentrum fuer Technik GmbH (AZT), Ismaning (Germany)

    2002-07-01

    In the cooling system of the Grohnde Nuclear Power Plant, heavy calcium carbonate precipitations have occurred in the turbine condenser and in the cooling tower since 1994. Those precipitations cause performance losses and high cleaning costs. Reasons for the intensified precipitation are apparently the biologically upgraded water quality of the river Weser, as well as specific operational conditions of the cooling system (partial recirculation of cooling water). It was noticed that the formation of calcium carbonate deposits occurred within a limited period of less than two weeks in May. The calcium precipitation was that strong during this time that it could not be stopped despite the immediate application of corundum cleaning balls in the tube cleaning system. (orig.) [German] Im Kuehlsystem des Kernkraftwerks Grohnde kommt es seit 1994 zu starken Kalkausfaellungen im Turbinenkondensator und im Kuehlturm, die Leistungsverluste und hohe Reinigungskosten verursachen. Das Auftreten der Kalkabscheidungen haengt offensichtlich mit der oekologisch verbesserten Wasserqualitaet der Weser und spezifischen Betriebsbedingungen des Kuehlsystems zusammen. Die Auswertung der Betriebsmessdaten des Kuehlwassers zeigt, dass eine der Ursachen der zeitlich begrenzten, extremen Kalkabscheidung im Kuehlsystem offenbar das Wachstum von Mikroorganismen in der Weser ist, aus der das Kuehlsystem gespeist wird. (orig.)

  2. 岭澳核电站二期LOFW+ATWS事故的氢气风险研究%Hydrogen Safety Analysis of Ling'ao Ⅱ NPP under LOFW+ATWS Accident

    Institute of Scientific and Technical Information of China (English)

    黄兴冠; 杨燕华; 傅孝良

    2011-01-01

    应用安全壳内氢气安全分析程序(GASFLOW)模拟了岭澳核电站二期在失去给水+未能紧急停堆的预计瞬变(LOFW+ATWS)事故下安全壳内氢气和水蒸汽的行为,对事故进程中氢气的风险进行了安全分析,特别是对氢气缓解系统的效果进行了评价.模拟结果说明,安全壳内温度与压力的变化与水蒸汽的喷放密切相关;水蒸汽在安全壳内会呈现一定的分层现象;泄压箱隔间与稳压器隔间在氢气释放峰值阶段可能发生火焰加速现象.%This paper applies the three dimensional CFD code GASFLOW to simulate the behavior of hydrogen and steam under hypothetical LOFW+ATWS accident. Analysis shows that the average temperature and pressure are dominated by the inventory of the steam released to the containment. The calculation results also show that during the hydrogen release time, the average recombine rate of the hydrogen mitigation system can reach 40g/s, but the flame acceleration criterion in the code assess that the flame acceleration may occur in quench tank compartment and pressurizer compartment during hydrogen release peak.

  3. 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.

  4. KWL Lingen nuclear plant. Technical annual report 2015; KWL Kernkraftwerk Lingen. Technischer Jahresbericht 2015

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-07-01

    The technical annual report 2015 on the Lingen nuclear plant covers the following issues: report on the segments operation, process engineering, safety engineering, licensing and supervising procedures, operational data, radiation protection, radioactive materials, and in-service inspections.

  5. 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 imp...

  6. Probability and consequences of severe reactor accidents. 60th year atw

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, Norman Carl [Massachusetts Institute of Technology (MIT), Cambridge, MA (United States). Dept. of Nuclear Engineering

    2015-06-15

    The study carried out on behalf of former USAEC (United States Atomic Energy Commission) led by Prof. Rasmussen and published in reworked form as WASH 1400 by the USNRC (United States Nuclear Regulatory Commission) in 1975, assessed in 3,300 pages the risks that can be deducted from severe accidents in nuclear power plants. The results, often quoted and criticised, were so far the most conclusive statements to this question. In his lecture at the reactor meeting in 1976, Prof. Rasmussen tried to trace back the conclusion of the results to the question: Is the use of larger nuclear power plants, in accordance to experiences and calculations so far, acceptable? His risk assessment, related to American power plants and cites, on behalf of the BMI is currently evaluated by the IRS together with the LRA on specific occurrences within the Federal Republic of Germany.

  7. Lifetime extension of nuclear power plants. Exclusive competence of the Bundestag?; Laufzeitverlaengerung fuer Kernkraftwerke. Entscheidung zwischen Bundestag und Bundesrat?

    Energy Technology Data Exchange (ETDEWEB)

    Scholz, Rupert [Muenchen Univ. (Germany). Inst. fuer Politik und oeffentliches Recht

    2010-05-15

    With the Act on the structured phase-out of the utilisation of nuclear energy for the commercial generation of electricity (Gesetz zur geordneten Beendigung der Kernenergienutzung zur gewerblichen Erzeugung von Elektrizitaet) of 22 April 2002 (Federal Gazette I p. 1351), the ''nuclear power phase-out'' was implemented into law. Ever since then, section 7 (1a) of the Atomic Energy Act (Atomgesetz - AtG) has provided that the authorisation to operate a nuclear power plant expires once the electricity volume for the respective installation as listed in Appendix 3, column 2 or the electricity volume derived from transfers has been produced. The coalition treaty of the current government factions provides for extending the operating periods of nuclear power plants. To this end, paragraphs 1a to 1d of section 7 AtG could be repealed, thus restoring the legal status prevailing prior to the ''phase-out''. As an alternative it would be conceivable to increase the values set forth in Appendix 3 for the energy volume quantity of a given installation accordingly. Both alternatives require an amendment of the Atomic Energy Act, over which the Deutsche Bundestag has exclusive competence. This is stated in the Grundgesetz (Constitution). Such a amendment would not require the consent of the Bundesrat, since the administrative tasks assigned to the Federal States (Laender) on behalf of the Federal Government pursuant to sec. 7, 24 (2) AtG would not be changing in a qualitative sense. Consequently, it would not constitute interference with the administrative powers of the Federal States from an organizational or procedural point of view. The quantitative change in the tasks to be performed by the Federal States on behalf of the Federal Government that would accompany an extension of the operating periods would not lead to a right of consent on the part of the Bundesrat pursuant to Art. 87c of the Grundgesetz. (orig.)

  8. VGH Mannheim: legitimacy of the decommissioning license for a nuclear power plant; VGH Mannheim: Rechtmaessigkeit der Stilllegungsgenehmigung fuer ein Kernkraftwerk

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2015-03-16

    The contribution describes the details of the court (VGH) decision on the legitimacy of the decommissioning license for the NPP Obrigheim. Inhabitants of the neighborhood (3 to 4.5 km distance from the NPP) are suspect hazards for life, health and property due to the dismantling of the nuclear power plant in case of an accident during the licensed measures or a terroristic attack with radioactive matter release.

  9. Harmonisation of licensing processes for decommissioning. Options and limitations; Genehmigungsverfahren fuer die Stilllegung der deutschen Kernkraftwerke. Konvoi oder Kakophonie?

    Energy Technology Data Exchange (ETDEWEB)

    Raetzke, Christian

    2016-03-15

    The shutdown of eight reactors in Germany in the wake of Fukushima 2011 and the scheduled phase-out of the remaining units in several steps ending 2022 has obviously triggered a wave of applications for decommissioning and dismantling licences. It would seem natural to strive for a harmonised handling of these processes, analogous to the 'convoi' concept which was successfully employed for licensing and construction of the three most recent German NPPs in the 1980s. However, a comparative analysis shows that the motivation of all players is much different from that of earlier times and that harmonisation of licensing processes for dismantling is not as crucial for operators, authorities and technical support organisations as it was for construction.

  10. 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)

  11. 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.)

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

    Science.gov (United States)

    1998-05-01

    surface and the renewal of water ; the estimation is true. . the degree of accessibility to the earth must be located close by (\\0602 miles) of cages...mathematical system theory provides a framework for representing and studying dynamical systems. The apparition of a large discrete event models coming from...with : • j 1 = Degree of protection of the site; • j2 = Wealth faunistique / floristique of the site; • j3 = Surface / depth of the water plan

  13. 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

    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...

  14. Nuclear power: on line; Kernenergie Online

    Energy Technology Data Exchange (ETDEWEB)

    Thieme, Christian [atw Redaktion, Hattingen (Germany)

    2011-04-15

    Presentation of these contents in the World Wide Web (WWW): Joint Research Centre (JRC), Institute for Transuranium Elements (ITU) - itu.jrc.ec.europa.eu Kernkraftwerk Gundremmingen (Germany) - www.kkw-gundremmingen.de Canadian Nuclear Association (Canada) - www.cna.ca Kernkraftwerk Krsko (Slovenia) - www.nek.si. (orig.)

  15. 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.

  16. Tools and tool application for the dismantling of the nuclear power plant Brennilis in France; Werkzeuge und Werkzeugeinsatz fuer den Rueckbau des Kernkraftwerks Brennilis Frankreich

    Energy Technology Data Exchange (ETDEWEB)

    Bienia, Harald; Welbers, Philipp; Krueger, Peter; Noll, Thomas [NUKEM Technologies GmbH, Alzenau (Germany)

    2012-11-01

    The EL-4 reactor in the NPP Brennilis in France is a CO2 cooled heavy water moderated test reactor with net power of 70 MW, the reactor started operation in 1967 and was decommissioned in 1985. Due to the construction features it was not necessary to enter the reactor area during operation, therefore the reactor pressure vessel and the surrounding piping systems are built in a very compact way. The dismantling procedures are therefore different from German BWR or PWR systems, the remote cutting and handling tools have to be adapted to the different features. Because of the high local dosage rate in the reactor hall it is also necessary to perform the erection of the dismantling equipment by robot systems. For cutting of the piping system a new plasma cutting technique, the hot wire method will be used. Other mechanical cutting techniques have to be used for instance for zircaloy containing components due to fire prevention purposes. The required time for tool and manipulator changes, including wearing part replacements constitute a significant part of the dismantling schedule. The suction/exhaust system for radioactive dust removal allowed a reduction of the total personal dose by one third of the allowed dose.

  17. Added value by rule IEC61850. Modernizing the electrical protection of Gundremmingen nuclear power plant; Mehrwert durch die Norm IEC 61850. Erneuerung des Blockschutzes im Kernkraftwerk Gundremmingen

    Energy Technology Data Exchange (ETDEWEB)

    Hoetzel, A.; Willems, D. [RWE Rhein-Ruhr AG, Brauweiler (Germany); Maier, K.L. [Kernkraftwerk Gundremmingen GmbH (Germany); Herrmann, H.J.; Einsiedler, G. [Siemens Power Transmission and Distribution (PTD), Nuernberg (Germany)

    2006-07-01

    After many years in operation the large power plant generating units B and C at Gundremmingen nuclear power plant are due for inspection and maintenance, which also requires modernizing the electrical protection. Unlike the construction of new power plants, additional constraints apply to modernization in existing plants. The new solution has to fit as seamlessly as possible into the existing units, such as signaling systems with their multitude of signaling contacts and printers, or the connection to the power plant automation system. Apart from purely technical requirements, economic factors such as short standstill times, limited budgets or phased conversions also influence the choice of a suitable solution. Planning, construction and commissioning of the electrical generating unit protection was implemented by the Secondary Systems Technology Center, a technical department of RWE-Rhein-Ruhr Netzservice GmbH, in coordination with the operator. (orig.)

  18. Successful implementation of ageing management exemplified at the cooling tower of Emsland nuclear power plant; Erfolgreiche Umsetzung von Alterungsmanagement am Beispiel Kuehlturm des Kernkraftwerkes Emsland

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Alexander [Hochtief Solutions AG, Consult IKS Energy, Frankfurt am Main (Germany). Design Kraftwerke; Dueweling, Carsten [Kernkraftwerke Lippe-Ems GmbH, Lingen (Germany). Abschnitt Bautechnik

    2013-07-15

    The present paper describes the successful implementation of the restoration of water-distribution channels at the cooling tower of the Emsland nuclear power plant under the aspect of ageing management. The main challenge of aging management is the determination of potential aging mechanism and to avoid systematically and effectively their damaging influences. In the course of the annual site inspections abnormalities at the lower side of the water-distribution channels of the cooling tower were detected, analysed, and repaired. The extraordinary high chlorine equivalent of the cooling water was identified as main reason of the damages located. Due to extensive infiltration into the concrete structure, chloride-induced corrosion generates a volume expansion of the reinforcement and thereby to a blast off of the concrete covering. According to the restoration concept, the damaged concrete was removed by maximum pressure water jet blasting; where necessary the reinforcement was retrofitted and a layered concrete substitution was applied by synthetic cement mortar. The realised procedures conserve the load bearing reinforcement only for a certain period, because the permanent chloride infiltration could not be stopped. Therefore, the structure has to be monitored permanently. (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. 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.

  1. Periodical safety review of the Goesgen-Daeniken nuclear power plant. Summary, results and evaluation; Periodische Sicherheitsueberpruefung fuer das Kernkraftwerk Goesgen-Daeniken. Zusammenfassung, Ergebnisse und Bewertung

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-15

    The Goesgen nuclear power plant (KKG) received its operational licence on September 9, 1978. The operational start-up of the plant went on into the year 1979, but there was a short interruption because of the accident in the Three Mile Island reactor on March 28, 1979. In May 1985 KKG submitted a request for raising the thermal reactor power from the then 2808 MW to 3002 MW. Based on the examination by the Federal Agency for the Safety of Nuclear Installations (HSK), the Swiss Federal Council granted the licence in two steps: in December 1985 for raising the thermal power to 2900 MW, and, in April 1992, to 3002 MW. The licence for the second step was given under the condition that some more experience was to be gained concerning the fuel rod cladding under higher loading. As part of the yearly re-licensing on restart after fuel assembly reloading, HSK confirmed that the plant status conformed to the legal requirements. In November 1986, HSK asked all Swiss nuclear power plant managers to state their opinions on proposed measures concerning severe accidents. Some of the measures were already in discussion; the Chernobyl accident on April 26, 1986, accelerated their implementation and was also a reason for the introduction of the measures against severe accidents. In this context, KKG carried out a risk study which led to the installation of a filtered pressure release system for the containment. Another consequence of the Chernobyl accident was the introduction of technical Periodical Safety Reviews (PSR) for all operating nuclear power plants. Central points of the PSR are: a) comparison with the continuously improving state-of-the-art of science and technology concerning safety precautions; b) a systematic evaluation of operating experience and plant status; c) the taking into account of probabilistic safety analyses in the overall evaluation of the plant. Within the framework of the examination of the overall plant, HSK also checks how its requirements concerning plant safety and radiation protection are taken into account. Even if the plant manager considers the guarantee of plant safety as his duty, an overall investigation by the authorities makes sense because it also looks into rare accident scenarios for which there are, of course, no actual working experience and which can only be considered within the framework of extended plant examinations. The PSRs on the Swiss nuclear power plants therefore complement the continuous control activities of the HSK; they are carried out about every 10 years. For KKG the PSR process was initiated by a letter from the HSK in February 1994. The areas to be considered were: a) examination of design and fulfilment of technical safety systems and comparison with the actual state-of-the-art of science and technology; b) evaluation of operational experience; c) review of the technical precautions against severe accidents including the preparation of emergency measures; d) review of the emergency organisation; e) examination of the plant protection against radioactivity; f) future dismantling at the end of operational life and disposal of the radioactive wastes; g) evaluation of accident analyses and of the KKG probabilistic safety analysis; h) review of plant organisation and plant management. The examination confirmed that, at KKG, there are very many technical safety precautions. KKG operational experience is good, the results show a high degree of operational availability and a very low number of incidental shut-downs. In international comparison the collective doses of the staff are low and the release of radioactive materials to the environment is negligible; on this account KKG is one of the world's best plants operating pressurised water reactors. Up to now the examinations have not brought any ageing deterioration to light concerning the status of safety-relevant components or ducts

  2. 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

    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...... a chemical solution deposition route....

  3. 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

    frequent in this substrate, including both coherent and incoherent twin boundary segments as well as non-twin R3 boundaries formed between twins and non-matrix neighbors of the cube texture component. A strong correlation between the boundary type and the average depth of grain boundary grooves...... was observed in this material. The smallest average groove depth was obtained for coherent twin boundaries, followed by low angle boundaries. Significantly greater average groove depths were found for the other boundary types. A similar correlation was also observed between the boundary type and the average...

  4. 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

    High performance long-length coated conductors fabricated using various techniques have attracted a lot of interest recently. In this work, a reel-to-reel design for depositing double-sided coatings on long-length flexible metallic tapes via a chemical solution method is proposed and realized....... 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...... 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...

  5. 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

  6. Project 'WINDBANK mittleres Aaretal' - Analysis, Diagnosis and Forecast of Wind Fields around the Nuclear Power Plant Goesgen; Projekt 'WINDBANK mittleres Aaretal' - Analyse, Diagnose und Prognose der Windverhaeltnisse um das Kernkraftwerk Goesgen

    Energy Technology Data Exchange (ETDEWEB)

    Graber, W.K.; Tinguely, M

    2002-07-01

    An emergency decision support system for accidental releases of radioactivity into the atmosphere providing regional wind field information is presented. This system is based on intensive meteorological field campaigns each lasting 3-4 months in the regions around the Swiss nuclear power plants. The wind data from temporary and permanent stations are analysed to evaluate the typical wind field patterns occurring in these regions. A cluster analysis for these data-sets lead to 12 different wind field classes with a high separation quality. In the present report, it is demonstrated that an on-line acquisition of meteorological data from existing permanent stations is enough to diagnose the recent wind field class in a region with a radius of 25 km around the nuclear power station of Goesgen with a probability of 95% to hit the correct class. Furthermore, a method is presented to use a high resolution weather prediction model to forecast the future wind field classes. An average probability of 76% to hit the correct class for a forecast time of 24 hours is evaluated. Finally, a method for parameterization of turbulence providing input for dispersion models from standard meteorological online data is presented. (author)

  7. Flaws in NPP Unterweser steam generators' feedwater nozzles - Detection, root causes, restoration and preventive measures; Befunde an den Speisewasserstutzen von Dampferzeugern des Kernkraftwerkes Unterweser - Detektierung, Ursachenklaerung, Sanierung, Abhilfemassnahmen

    Energy Technology Data Exchange (ETDEWEB)

    Kohlpaintner, W.; Nowak, E.; Schmidbauer, J.; Wachter, O. [E.ON Kernkraft, Hannover (Germany); Neumann, J.; Voskamp, R. [E.ON Kernkraftwerk Unterweser, Stadland (Germany); Wesseling, U. [Framatome ANP GmbH, Erlangen (Germany)

    2003-07-01

    During recurrent inspections, damage was found in the feedwater nozzle of the KKU DE10 steam generator. The other three steam generators were tested as well, with damage found in two further nozzles (DE 20/40). No damage was found in DE30. The findings were validated by further nondestructive tests. Standstill corrosion in the past was identified as the cause of damage. The damaged nozzles were exchanged, and the KKU is back in operation. As a preventive measure for the future, optimisations in the mode of operation were made as well (reduced temperature loads, wet conservation). (orig.) [German] Im Rahmen wiederkehrender zerstoerungsfreier Pruefungen (Ultraschall(US)-Standardpruefung) wurde am Speisewasserstutzen des KKU-Dampferzeugers DE10 ein Befund an der Innenoberflaeche des Stutzenrohres an der 6 -Position im Uebergangsradius Stutzenrohr/Thermosleeve festgestellt. Eine vorsorgliche Ueberpruefung der Speisewasserstutzen an den uebrigen drei DE ergab Anzeigen an zwei weiteren Stutzen (DE 20/40). In dem Speisewasserstutzen des Dampferzeugers DE30 wurden keine Anzeigen festgestellt. Weitere, zur Absicherung durchgefuehrte zerstoerungsfreie Pruefungen (US-Analysetechnik / Durchstrahlungspruefung) bestaetigten die mit der US-Standardpruefung aufgefundenen Anzeigen. Die US-Analysetechnik ergab im Vergleich zur US-Standardprueftechnik jedoch groessere Anzeigentiefen, so dass zur Ursachenklaerung weitergehende Schadensuntersuchungen durchgefuehrt wurden. Hierzu wurde der befundbehaftete Bereich eines DE-Stutzens herausgetrennt und werkstofftechnischen Untersuchungen unterzogen. Die Ergebnisse der durchgefuehrten Untersuchungen zeigen, dass der festgestellte Befund auf in der Vergangenheit aufgetretene Phaenomene der Stillstandskorrosion in Verbindung mit lokalen Spannungserhoehungen zurueckzufuehren ist. Es wurde entschieden, den befundfreien Zustand bei den betroffenen drei DE-Speisewasserstutzen wiederherzustellen. Hierzu wurden die Stutzenbereiche mit Schaedigung durch geschmiedete Ringe gleichen Werkstoffes ausgetauscht. Die Reparaturmassnahmen sind zwischenzeitlich abgeschlossen. Die Anlage KKU ist wieder in Betrieb. Zur Minimierung der Stutzenbelastungen im weiteren Anlagenbetrieb wurden vorsorglich noch Optimierungen an der Betriebsweise (Verringerung Temperaturbelastung, Nasskonservierung) vorgenommen. (orig.)

  8. 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.

  9. 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.

  10. Analysis of the EU stress test results for the NPP Fessenheim and Beznau. Pt. 2. Beznau; Analyse der Ergebnisse des EU-Stresstest der Kernkraftwerke Fessenheim und Beznau. T. 2. Beznau

    Energy Technology Data Exchange (ETDEWEB)

    Brettner, Mathias [Physikerbuero Bremen (Germany); Pistner, Christoph; Kurth, Stephan [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany)

    2012-10-11

    As a consequence of the reactor accidents in Fukushima Daiichi the safety status of nuclear power plants was performed by national and international surveillance processes. In Germany the safety status of the NPP was testes by the reactor safety commission, an expert commission of Baden-Wuerttemberg, an expert commission of Bavaria and the EU stress test. The evaluation criteria based on national and international surveillance processes were focused on earthquakes, flooding and postulated failures of the electricity supply - station blackout and long-lasting failure of the emergency power supply. In Germany extended requirements included the electricity supply and the emergency cooling water supply. The authors identify essential safety relevant systems in the NPP Beznau, including technical systems, energy supply, plant-internal emergency measures, and discuss specific Swiss requirements. The evaluation of the EU stress test for the NPP Fessenheim covers the issues earthquake, flooding, spent fuel element pool, electricity supply, cooling water supply and identification of further safety relevant deficiencies.

  11. Analysis of the EU stress test results for the NPP Fessenheim and Beznau. Pt. 1. Fessenheim; Analyse der Ergebnisse des EU-Stresstest der Kernkraftwerke Fessenheim und Beznau. T. 1. Fessenheim

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Kueppers, Christian; Kurth, Stephan; Mohr, Simone [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany); Brettner, Mathias [Physikerbuero Bremen (Germany)

    2012-10-11

    As a consequence of the reactor accidents in Fukushima Daiichi the safety status of nuclear power plants was performed by national and international surveillance processes. In Germany the safety status of the NPP was testes by the reactor safety commission, an expert commission of Baden-Wuerttemberg, an expert commission of Bavaria and the EU stress test. The evaluation criteria based on national and international surveillance processes were focused on earthquakes, flooding and postulated failures of the electricity supply - station blackout and long-lasting failure of the emergency power supply. In Germany extended requirements included the electricity supply and the emergency cooling water supply. The authors identify essential safety relevant systems in the NPP Fessenheim. The evaluation of the EU stress test for the NPP Fessenheim covers the issues earthquake, flooding, spent fuel element pool, electricity supply, cooling water supply and identification of further safety relevant deficiencies.

  12. Reserves for nuclear power plant decommissioning and radwaste disposal in Germany. An analysis and evaluation from the angle of energy policy; Energiewirtschaftliche Bewertung der Rueckstellungen fuer die Entsorgung der deutschen Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Buerger, V. [comp.

    1998-12-31

    The study, which is the first of its kind in Germany, presents a comprehensive survey of total reserves set up by the German nuclear industry for liabilities and costs for nuclear power plant decommissioning and resulting radwaste disposal, which is a legal and foreseeable responsibility but uncertain in amount. The study looks into the various ways the earmarked money was invested and analyses the funds with respect to their efficiency and reliability to provide financial security for the given tasks and purpose. The question put in this context is: Are the reserves set up so far in line with official cost estimates, i.e. will they cover estimated costs, or do they even exceed the estimated amounts? The conclusions drawn and explained in this document are: The reserves for nuclear decommissioning have been used by the nuclear power plant operators and electricity companies as a significant capital source. Some of the capital accrued is being increasingly used at present to cover expenses arising for restructuring of business and diversification into new business segments of interest in the open national and European electricity markets. Companies such as RWE, Preussen Elektra, and Bayernwerk, which until deregulation of the energy sector were just power supply companies, have been transformed into conglomerate companies and international players in the markets, like RWE Holding, VEBA, and VIAG. It can be safely assumed that the companies would not have been able to reach the important positions they currently hold in the German economy without tapping the reserves for nuclear decommissioning. (orig./CB) 27 refs. [Deutsch] Die Studie gibt erstmals einen vollstaendigen Ueberblick ueber die Summe der in Deutschland gebildeten Rueckstellungen im Kernenergiebereich. Sie geht der Frage nach, wie diese Gelder angelegt sind und ob die praktizierten Anlageformen dem hohen Sicherheitsanspruch entsprechen, den die Gesellschaft an die finanziellen Ressourcen zur Bewaeltigung eines derart gravierenden Problems stellen muss. Weiter wird untersucht, ob die absolute Hoehe der gebildeten Rueckstellungen mit den offiziell diskutierten Kostenschaetzungen in Einklang steht; ob also ausreichend Gelder zurueckgestellt werden oder ob die Rueckstellungen gar zu hoch sind. Ein weiterer wichtiger Aspekt wird hierbei deutlich: Die Rueckstellungen im Kernenergiebereich sind von ihrer absoluten Hoehe her nicht mit Rueckstellungen zu anderen Zwecken z.B. fuer Pensionszahlungen, zu vergleichen. Sie haben fuer die rueckstellungsbildenden Unternehmen eine enorme Bedeutung als Kapitalquelle entwickelt. Dieses Kapital spielte und spielt immer noch eine erhebliche Rolle bei der Umwandlung von frueher vorwiegend auf den Stromsektor orientierten Unternehmen wie RWE, PreussenElektra und Bayernwerk zu modernen, international agierenden Mischkonzernen wie der RWE Holding, VEBA und VIAG. Es darf angenommen werden, dass die herausragende Rolle dieser Konzerne in der deutschen Volkswirtschaft ohne dieses Finanzpolster nicht in gleichem Umfang haette entwickelt werden koennen. (orig.)

  13. Reserves for nuclear power plant decommissioning and radwaste disposal in Germany. An analysis and evaluation from the angle of energy policy. Energiewirtschaftliche Bewertung der Rueckstellungen fuer die Entsorgung der deutschen Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Buerger, V. (comp.)

    1998-01-01

    The study, which is the first of its kind in Germany, presents a comprehensive survey of total reserves set up by the German nuclear industry for liabilities and costs for nuclear power plant decommissioning and resulting radwaste disposal, which is a legal and foreseeable responsibility but uncertain in amount. The study looks into the various ways the earmarked money was invested and analyses the funds with respect to their efficiency and reliability to provide financial security for the given tasks and purpose. The question put in this context is: Are the reserves set up so far in line with official cost estimates, i.e. will they cover estimated costs, or do they even exceed the estimated amounts The conclusions drawn and explained in this document are: The reserves for nuclear decommissioning have been used by the nuclear power plant operators and electricity companies as a significant capital source. Some of the capital accrued is being increasingly used at present to cover expenses arising for restructuring of business and diversification into new business segments of interest in the open national and European electricity markets. Companies such as RWE, Preussen Elektra, and Bayernwerk, which until deregulation of the energy sector were just power supply companies, have been transformed into conglomerate companies and international players in the markets, like RWE Holding, VEBA, and VIAG. It can be safely assumed that the companies would not have been able to reach the important positions they currently hold in the German economy without tapping the reserves for nuclear decommissioning. (orig./CB) 27 refs.

  14. How does react power price on a possible lifetime extension for power plants? Nuclear power, power prices and power market models; Wie reagiert der Strompreis auf eine moegliche Verlaengerung der Laufzeiten fuer Kernkraftwerke? Kernkraft, Strompreis und Strommarktmodelle

    Energy Technology Data Exchange (ETDEWEB)

    Nestle, Uwe [Buendnis 90/Die Gruenen, Berlin (Germany). Bundesarbeitsgemeinschaft Energie

    2010-08-23

    Extending the life of the nuclear power plants currently operated in Germany is being discussed in the light of a more likely change in government for a Christian Democrat/Liberal coalition. The reason cited most frequently is the impossibility to meet the objectives of climate protection without raising further the price of electricity if the life of nuclear power plants cannot be extended. The question to be looked into is that of the legal pre-requisites to be established in Germany in order for the existing nuclear power plants to be operated for longer periods of time. So in this contribution some discussion is done wether a possible lifetime extension of nuclear power plants will react on power prices.(GL)

  15. The German-German history of the nuclear power plant Greifswald. Nuclear power between east and west. 2. ed.; Die deutsch-deutsche Geschichte des Kernkraftwerkes Greifswald. Atomenergie zwischen Ost und West

    Energy Technology Data Exchange (ETDEWEB)

    Hoegselius, Per [Technische Hochschule Stockholm (Sweden). Bereich Gesellschaft, Wissenschaft und Technik

    2015-07-01

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

  16. 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.

  17. Project 'Windbank upper Aare Valley': Classification, diagnosis and prognosis of wind fields in the region of the Swiss nuclear power plant Muehleberg; Projekt 'Windbank oberes Aaretal': Klassifizierung, Diagnose und Prognose von Windfeldern in der Region des Kernkraftwerkes Muehleberg

    Energy Technology Data Exchange (ETDEWEB)

    Graber, W.K.; Tinguely, M

    1999-10-01

    In the framework of the project 'Windbank', wind field patterns in an area of 50 by 50 km{sup 2} in the Swiss Plateau around the nuclear power plant Muehleberg between the Alps and the Jura were measured with 22 temporary meteorological stations and 2 SODARs during four months in 1997. Hourly averages from this high resolution network were combined with meteorological information from routine stations and from a weather prediction model. This data-set comprises all available parameters influencing the complex wind flow in the investigated area between the Alps and the Jura. A cluster analysis for this data-set leads to 12 classes with a high separation quality. It is demonstrated, that an on-line acquisition of meteorological data from routine stations and from a weather prediction model can be used to diagnose the recent wind field class with a probability of 96 % to hit the correct wind field class. This diagnosis reveals wind fields with a very high spatial resolution in a very short time. Consequently, it is useful as a contribution to a decision support system for safety management after accidental releases of nuclear or chemical air pollutants. Further, a method is outlined to use the weather prediction model to forecast the wind field class. An average probability of 79 % to hit the correct wind field classes for a forecast time of up to 25 hours is evaluated. (author) [German] Im Rahmen des Projektes 'WINDBANK oberes Aaretal', einem Auftrag des Bundesamtes fuer Energie (BfE) der Schweiz an das Paul Scherrer Institut, wurden im Gebiet des oberen Aaretales im zweiten Halbjahr 1997 Windmessungen durchgefuehrt. Dabei kamen zusaetzlich zu den Daten von Routinestationen noch 22 PSI-Stationen und 2 SODARs fuer begrenzte Zeit zum Einsatz. Basierend auf diesen Daten wird im vorliegenden Bericht in Anlehnung an das Vorlaeuferprojekt 'WINDBANK unteres Aaretal' eine Clusteranalyse dargestellt, um fuer die Gegend 12 typische Windfelder zu erarbeiten. Ausgehend von den mittleren Windwerten fuer jede Klasse werden anschliessend 3-dimensionale Windfelder auf einem regelmaessigen Gitter im Umfeld von 50x50 km berechnet. Diese Windfelder bilden die Grundlage, um typische Dispersionsszenarien fuer eine hypothetische Quelle zu berechnen. Weiter wird eine Methode zur Bestimmung eines aktuellen Windfeldes aufgrund der Routine-Stationen (ohne PSI-Stationen und SODARs) erarbeitet, die in 96 % aller Faelle das richtige Windfeld diagnostiziert. Damit kann eine Windfelddiagnose im on-line-Betrieb realisiert werden. Zur optimalen Nutzung der Windfeldklassierung als Planungsinstrument fuer Massnahmen bei Reaktorstoerfaellen wird eine Methode dargestellt, die eine Prognose der in den naechsten 24 Stunden zu erwartenden Windfeldklassen erlaubt. Die Methode stuetzt sich auf das operationell eingesetzte 'Schweizer Modell' der Schweizerischen Meteorologischen Anstalt und kann ebenfalls in einen on-line-Betrieb der Windfeldprognose eingebunden werden. Die Qualitaet der Prognose der Windfeldklassen wird mit 79 % abgeschaetzt. (author)

  18. 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.

  19. Application of ultrasonic array transducers and robotics for external inspection of reactor pressure vessels - practical experience from refuelling operations at the Kruemmel NPP in 2000; Einsatz von Ultraschall-Gruppenstrahlerprueftechniken und neuer Robotik bei der Aussenpruefung von Reaktor-Druckbehaeltern - Einsatzerfahrungen im Kernkraftwerk Kruemmel im Brennelementwechsel 2000

    Energy Technology Data Exchange (ETDEWEB)

    Eggers, H.; Hein, E.; Zwahr, B. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany). Abt. Oeffentlichkeitsarbeit; Rathgeb, W.; Schirner, G. [Siemens NDT, Erlangen (Germany)

    2000-07-01

    The performance of a novel testing robot of Siemens NDT and novel testing techniques for inspection of a welded joint for longitudinal and transverse defects is described in great detail. (orig./CB) [German] Im Rahmen einer Pruefgeraetelieferung ist HEW und Siemens NDT uebereingekommen, die gesamte Prueftechnik fuer die Siedewasseraussenpruefung effizienter zu gestalten. Ziel war, mit einem voellig neuen Pruefroboter und neuer Prueftechnik eine Schweissnaht in einem Prueflauf vollstaendig auf Laengs- und Querfehler zu pruefen. Dabei sollten die Umbaumassnahmen zwischen den einzelnen Pruefbereichen, wie den Rundnaehten und den Stutzeneinschweissnaehten, sowie der Stutzenkante entfallen. Die Prueftechnik wurde auf eine moderne Gruppenstrahlerprueftechnik umgestellt. Die dabei erreichbare Pruefdichte sollte zumindest so, wie mit herkoemmlichen Pruefsystemen oder dichter sein. Grundlegende Untersuchungen wurden zur Tandemtechnik angestellt, um mit einer modifizierten Tandemtechnik, bestehend aus zwei Gruppenstrahlerpruefkoepfen, die Wanddicke des SWR-Reaktordruckbehaelters mit maximal fuenf Tandemzonen zu pruefen. (orig.)

  20. 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.)

  1. 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.

  2. Accelerator-driven destruction of long-lived radioactive waste and energy production

    Energy Technology Data Exchange (ETDEWEB)

    Schriber, S.O.

    1997-12-31

    Nuclear waste management involves many issues. ATW is an option that can assist a repository by enhancing its capability and thereby assist nuclear waste management. Technology advances and the recent release of liquid metal coolant information from Russia has had an enormous impact on the viability of an ATW system. It now appears economic with many repository enhancing attributes. In time, an ATW option added to present repository activities will provide the public with a nuclear fuel cycle that is acceptable from economic and environmental points of view.

  3. Process Performance Measurement in a Maintenance Process; Prozess Performance Measurement im Auftragsbearbeitungsprozess

    Energy Technology Data Exchange (ETDEWEB)

    Kronz, A. [IDS Scheer AG, Saarbruecken (Germany); Ramler, K. [Eon Kernkraft GmbH (Germany). Kernkraftwerk Unterweser; Renner, A. [Horvath und Partner GmbH (Germany)

    2002-07-01

    Collecting and analysing of process oriented performance indicators is a prerequisite to a comprehensive process management. In order to realize continuous process optimization, performance gaps within the processes of companies are being detected. Thus, an adequate IT-solution like ARIS-Process Performance Manager is needed. The following presents the results of a project at the nuclear power plant Eon Kernkraftwerk Unterweser, which was carried out by Horvath and Partners and IDS Scheer. (orig.) [German] Die Erhebung und Analyse leistungsbezogener Prozesskennzahlen ist Voraussetzung fuer ein ganzheitliches Prozessmanagement. Nur so laesst sich eine durchgaengige und kontinuierliche Prozessoptimierung realisieren. Zudem werden Schwachstellen in Unternehmensprozessen aufgedeckt. Voraussetzung ist eine geeignete IT-Loesung wie der Aris-Process-Performance-Manager. Vorgestellt werden die Ergebnisse eines von Horvath and Partners und IDS Scheer gemeinsam durchgefuehrten Projekts bei dem Eon-Kernkraftwerk Unterweser. (orig.)

  4. WTZ Russland - Fluenzberechnungen für Voreilproben beim WWER-440

    OpenAIRE

    Konheiser, Jörg; Grahn, Alexander

    2014-01-01

    Der Reaktordruckbehälter (RDB) zählt zu den nicht auswechselbaren Komponenten eines Kernkraftwerkes (KKW). Durch die hohen Neutronen- und Gammaflüsse ist er beschleunigten Alterungsprozessen unterworfen, welche die Lebensdauer eines KKW bestimmen könnten. So haben neben der chemischen Zusammensetzung des RDB-Stahls vor allem die Strahlungsparameter (Neutronen- und Gammafluenzen und deren Spektren) Auswirkungen auf die Versprödungseigenschaften des RDB. Für einen sicheren Betrieb eines KKW ...

  5. 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 2016-06-30 (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...

  6. 46{sup th} annual meeting on nuclear technology (AMNT 2015). Key topics / Outstanding know-how and sustainable innovations enhanced safety and operation excellence

    Energy Technology Data Exchange (ETDEWEB)

    Lamm, Matthias [AREVA GmbH, Erlangen (Germany). R and D; Hollands, Thorsten [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Reactor Safety Research Div.

    2016-01-15

    Summary report on the Technical Sessions ''Know-how, New Build and Innovations'' and ''Operation and Safety of Nuclear Installations, Fuel SA: WASA-BOSS + CESAM'' of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 12 (2015) and will be covered in further issues of atw.

  7. Interdecadal change of atmospheric stationary waves and North China drought

    Institute of Scientific and Technical Information of China (English)

    Dai Xin-Gang; Fu Cong-Bin; Wang Ping

    2005-01-01

    The inderdecadal change of atmospheric stationary waves (ATW) has been investigated for the two periods 1956-77 and 1978-99. The trough of ATW in the middle and low layer of the troposphere over the Asian continent has experienced a significant weakening during the past two decades, which exerts a great influence on the North China climate. The ATW in 200 hPa has also exhibited some changes since 1977, as a stationary ridge appeared over the northwestern China while a stationary trough appeared above North China. This leads to an increasing of the upward motion above northwestern China and a decreasing above North China. A west-east section of the stationary waves at 40°N shows that the ATW above North China tilted westward for the period 1956-77, but was almost upright during 1978-99. The composite analysis confirms that the climate mean ATW pattern after 1977 is similar to the dry pattern for North China, while the rainy pattern is similar to that before 1977. In consequence, the North China drought is partly due to the interdecadal change of the ATW over boreal Asia in the recent two decades.

  8. Planning the research and development necessary for accelerator transmutation of waste, leading to integrated proof of performance testing

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, D.R.; Pasamehmetoglu, K.; Finck, P.; Pitcher, E.; Khalil, H.; Todosow, M.; Hill, R.; Van Tuyle, G.; Laidler, J.; Crawford, D.; Thomas, K. [Los Alamos National Lab., NM (United States)]|[Argonne National Lab., IL (United States)]|[Brookhaven National Lab., Upton, NY (United States)

    2001-07-01

    The Research and Development (R and D) Plan for the Accelerator Transmutation of Waste (ATW) Program has been developed for the Department of Energy, Office of Nuclear Energy (DOE/NE) to serve as a focus and progressional guide in developing critical transmutation technologies. It is intended that the Plan will serve as a logical reference considering all elements of an integrated accelerator-driven transmutation system, and will maximize the use of resources by identifying and prioritizing research, design, development and trade activities. The R and D Plan provides a structured framework for identifying and prioritizing activities leading to technically-justifiable integrated Proof of Performance testing within ten years and ultimate demonstration of Accelerator Transmutation of Waste (ATW). The Plan builds from the decision objectives specified for ATW, utilizes informational input from the ATW Roadmap and programmatic System Point Design efforts, and employs the knowledge and expertise provided by professionals familiar with ATW technologies. With the firm intent of understanding what, why and when information is needed, including critical interfaces, the Plan then develops a progressional strategy for developing ATW technologies with the use of a Technology Readiness Level (TRL) scale. The TRL approach is first used to develop a comprehensive, yet generic, listing of experimental, analytical and trade study activities critical to developing ATW technologies. Technology-specific and concept-specific aspects are then laid over the generic mapping to gage readiness levels. Prioritization criteria for reducing technical uncertainty, providing information to decision points, and levering off of international collaborations are then applied to focus analytical, experimental and trade activities. (author)

  9. 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.

  10. Relationship between T-wave amplitude and oxygen pulse in guinea pigs in hyperbaric helium and hydrogen.

    Science.gov (United States)

    Kayar, S R; Parker, E C; Aukhert, E O

    1998-09-01

    Diving is known to induce a change in the amplitude of the T wave (ATw) of electrocardiograms, but it is unknown whether this is linked to a change in cardiovascular performance. We analyzed ATw in guinea pigs at 10-60 atm and 25-36 degreesC, breathing 2% O2 in either helium (heliox; n = 10) or hydrogen (hydrox; n = 9) for 1 h at each pressure. Core temperature and electrocardiograms were detected by using implanted radiotelemeters. O2 consumption rate was measured by using gas chromatography. In a previous study (S. R. Kayar and E. C. Parker. J. Appl. Physiol. 82: 988-997, 1997), we analyzed the O2 pulse, i.e., the O2 consumption rate per heart beat, in the same animals. By multivariate regression analysis, we identified variables that were significant to O2 pulse: body surface area, chamber temperature, core temperature, and pressure. In this study, inclusion of ATw made a significantly better model with fewer variables. After normalizing for chamber temperature and pressure, the O2 pulse increased with increasing ATw in heliox (P = 0.001) but with decreasing ATw in hydrox (P pulse for animals breathing heliox vs. hydrox.

  11. 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.

  12. AMNT 2014. Key Topic: Fuel, decommissioning and disposal - report. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Seipolt, Thomas [NUKEM Technologies Engineering Services GmbH, Alzenau (Germany); Weber, Stefan [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Kock, Ingo [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) GmbH, Koeln (Germany)

    2015-02-15

    Summary report on the following Topical Sessions of the Key Topic 'Fuel, Decommissioning and Disposal' of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - From Pilot Project to an Industrial Service (Thomas Seipolt); - Radioactive Waste Management - Experiences with Interim and Final Storage (Stefan Weber and Ingo Kock). The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 and 12 (2014), 1 (2015) and will be covered in further issues of atw.

  13. 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.)

  14. Ecology and Molecular Genetic Studies of Marine Bacteria.

    Science.gov (United States)

    1988-01-31

    probes. ’p 5. Complete sequence analysis of the chi operon in pATW 501 and extrapolate the methodology to cloning of the urease gene. .: 6. The cloned...S. Nicklan and A.R. Coulson. 1977. DNA sequencing with chain-terminating inhibitors . Proc. Natl. Acad. Sci. USA, 74(12) :5463-5467. Somerville, C.C

  15. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key topic / Outstanding know-how and sustainable innovations

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Bereich Reaktorsicherheitsforschung

    2016-07-15

    Summary report on the Technical Session: ''Reactor Physics, Thermo- and Fluid-Dynamics'' of the 47th 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. 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.

  17. 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.

  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 con

  19. 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.

  20. 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

  1. Siting of nuclear power plants in accordance with the scenario of a ''fossil nuclear energy mix'' prepared by the CDU/CSU and FDP in the inquiry commission of the Bundestag on ''Sustainable energy supply in times of globalisation and deregulation''. Short study; Standortbestimmung fuer Kernkraftwerke gemaess dem von CDU/CSU und FDP definierten Szenario 'fossil-nuklearer Energiemix' der Enquete-Kommission 'Nachhaltige Energieversorgung unter den Bedingungen der Globalisierung und der Liberalisierung' des Deutschen Bundestages. Kurzstudie

    Energy Technology Data Exchange (ETDEWEB)

    Benik, G.; Gundelach, T.

    2002-08-28

    In effect the present study makes concrete proposals for the installation of a nuclear power plant capacity of 91.8 GW as determined by the CDU/CSU and FDP in the inquiry commission on ''Sustainable development''. Because of the infrastructural requirements and the large number of plants that would have to be planned in Germany alone these siting proposals must be regarded as an outline for the implementation of a nuclear climate protection strategy. Without extensive assessment work a short study such as this cannot of course replace the detailed planning that would be required for authorisation procedures. The primary benefit of the study is that it makes clear the true consequences of a nuclear climate protection policy that is now being discussed as a hypothetical possibility. [German] Die vorliegende Studie macht im Ergebnis entsprechend der Aufgabenstellung konkrete Vorschlaege fuer die Bereitstellung der im von CDU/CSU und FDP in der Enquete-Kommission ''Nachhaltige Entwicklung'' definierten Szenario ''Fossilnuklearer Energiemix'' ermittelten nuklearen Kraftwerksleistung von 91,8 GW Diese Standortvorschlaege muessen angesichts der infrastrukturellen Anforderungen und der grossen Zahl von zu planenden Anlagen allein innerhalb der Grenzen der Bundesrepublik Deutschland als Anhaltspunkte fuer die Umsetzung einer nuklearen Klimaschutzstrategie herangezogen werden. Dabei kann eine Kurzstudie selbstverstaendlich ohne umfaengliche Pruefung keine detaillierten Planungen ersetzen, welche im Rahmen entsprechender Genehmigungsverfahren erforderlich sind. Der Nutzen der Studie besteht in erster Linie darin, deutlich vor Augen zu fuehren, welche realen Konsequenzen eine zur Zeit noch hypothetisch gefuehrte Diskussion ueber den Einsatz der Kernkraft zum Klimaschutz hat. (orig.)

  2. A methodology for the analysis of a thermal-hydraulic phenomenon investigated in a test facility

    Energy Technology Data Exchange (ETDEWEB)

    D`Auria, F. [Dept. of Mechanical and Nuclear Constructions, Pisa Univ. (Italy); Faluomi, V. [Dept. of Mechanical and Nuclear Constructions, Pisa Univ. (Italy); Aksan, N. [Lab. for Thermal-Hydraulics, Paul Scherrer Inst., Villigen (Switzerland)

    1995-08-01

    A methodology for analysing non-homogeneous sets of experimental data for a selected phenomenon from separate effect test facilities and integral test facilities is presented in this paper. The critical heat flux from the validation matrices was chosen as the phenomenon to be studied; the results obtained in three test facilities are analysed. The method presented is applied for estimating the accuracy with which a thermalhydraulic transient code can predict the critical heat flux in an actual nuclear power plant. (orig.) [Deutsch] Gegenstand des Beitrags ist ein Verfahren zur Analyse ungleichartiger Datensaetze, die bei der experimentellen Untersuchung eines bestimmten thermohydraulischen Phaenomens in speziellen oder integralen Testeinrichtungen gewonnen wurden. Bei dem untersuchten Phaenomen handelt es sich hier um die kritische Waermestromdichte; experimentelle Daten aus drei Testeinrichtungen werden analysiert. Das Verfahren wird benutzt, um die Genauigkeit abzuschaetzen, mit der ein thermohydraulischer Rechencode zur Beschreibung von Uebergangszustaenden die kritische Waermestromdichte in einem Kernkraftwerk vorhersagen kann. (orig.)

  3. The Chernobyl reactor accident, ten years on. Teaching projects for mathematics instruction in interdisciplinary working groups; 10 Jahre nach Tschernobyl. Unterrichtsprojekte fuer den Mathematikunterricht in faecheruebergreifenden Kooperationen

    Energy Technology Data Exchange (ETDEWEB)

    Boer, H. [comp.; Delle, E. [comp.; Mies, K. [comp.; Warmeling, A. [comp.

    1996-10-01

    The booklet presents background information and addresses the following aspects: ionizing radiation and radiation effects; safety of German nuclear power plants; statistical evidence of radiation injuries; short-lived and long-lived ionizing radiation; radioactive waste; CO{sub 2} emissions as an argument in favour of nuclear power generation. The material presented is intended for use by a school project team interested in the subjects, or as a basis for collaborative, interdisciplinary teaching in working groups, and it offers information and problems for mathematics teaching. (HP) [Deutsch] Neben vielen Informationen behandelt die Broschuere: Strahlen und Strahlenwirkungen; Sicherheit deutscher Kernkraftwerke; statistischer Nachweis von Strahlenschaeden; Kurz- und Langfestigkeit der Strahlenbelastung; radioaktiver Abfall; CO{sub 2}-Problematik als Argument fuer die Kernenergie. Die Broschuere ist gedacht z.B. fuer eine Projektgruppe, einen Projekttag, fuer eine Lerngruppe in faecheruebergreifender Kooperation. Die Materialien sind ausgearbeitet fuer die Themembearbeitung im Mathematikunterricht mit Uebungsaufgaben. (HP)

  4. Fabrication of YBCO Coated Conductors on Biaxial Textured Metal Substrate by All-Sputtering

    Institute of Scientific and Technical Information of China (English)

    Xiao Han; Jing-Tan He; Jie Xiong; Bo-Wan Tao

    2008-01-01

    CeO2/YSZ/CeO2 buffer layers were prepared on biaxial textured Ni-5at.%W substrate by direct-current magnetron reactive sputtering with the optimum process. YBCO thin films were deposited on CeO2/YSZ/CeO2 buffered Ni-5at.%W substrate at temperature ranging from 500°C to 700°C by diode dc sputtering. By optimizing substrate temperature, pure c-axis oriented YBCO films were obtained. The microstructures of CeO2/YSZ/CeO2 buffer layers were characterized by X-ray diffraction. A smooth, dense and crack-free surface morphology was observed with scanning electron microscopy. The critical current density Jc about 0.75 MA/cm2 at 77 K was obtained.

  5. Morphological Atherosclerosis Calcification Distribution (MACD) Index is a Strong Predictor of Cardio-Vascular Death and Include Predictive Power of BMD

    DEFF Research Database (Denmark)

    Christiansen, Claus; Karsdal, Morten; Ganz, Melanie;

    followed for 8.3±0.3 years and CVD deaths were recorded. BMD and several aortic calcification markers were computed: number, morphology, distribution, from outlines of the calcified plaques in lumbar X-rays. These markers were compared to BMD, SCORE card, Framingham score, and the Aortic Calcification...... Severity score - AC24. AC24 adjusted by age, waist circumference, and triglyceride levels (ATW) predicted mortality in postmenopausal women (CVD p=0.03, All-cause p=0.006). The SCORE card and the Framingham score resulted in mortality odds ratios (MOR) of 5.0 and 5.2 - defining high risk as =6 and =18......, respectively. BMD and BMD adjusted for ATW was lower in the group of deceased than in survivors (pscores based on the calcification geometry provided highly significant predictions. The number of calcified deposits...

  6. Annual meeting on nuclear technology 2013. Pt. 2. Section reports

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, U.; Hoehne, Thomas [Forschungszentrum Dresden-Rossendorf (FZD) e.V., Dresden (Germany); Seidl, Marcus [E.ON Kernkraft GmbH, Hannover (Germany); Rossbach, Detlev [AREVA GmbH, Erlangen (Germany); Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Klute, Stefan [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany); Willmann, Frank [Toshiba International Europe Ltd., Uxbridge (United Kingdom)

    2013-10-15

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

  7. 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.)

  8. Naval Research Laboratory Major Facilities 2008

    Science.gov (United States)

    2008-10-01

    Mexico seawater throughout the year. The tropical climate is ideally suited for marine exposure testing. There is minimal climatic variation and a...TW magnetically insulated inductive voltage adder ( IVA ). Mercury is a focal point of research for several areas, including IVA power-flow research...nuclear weapons effects simulation, and particle-beam source and transport research for various applications. DESCRIPTION: Mercury is a 6-stage IVA . The

  9. 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.

  10. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key topic / Outstanding know-how and sustainable innovations

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

    2016-11-15

    Summary report on the Key Topic ''Outstanding Know-How and Sustainable Innovations'' Technical Session ''Reactor Physics, Thermo, and Fluid Dynamics'' of the 47th 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.

  11. 2011 annual meeting on nuclear technology. Topical sessions. Pt. 5; Jahrestagung Kerntechnik 2011. Fachsitzungsberichte. T. 5

    Energy Technology Data Exchange (ETDEWEB)

    Fazio, Concetta [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Nuclear Safety Research Programme

    2011-12-15

    Summary report on the Topical Session of the Annual Conference on Nuclear Technology held in Berlin, 17 to 19 May 2011: - Sodium Cooled Fast Reactors. The reports on the Topical Sessions: - CFD-Simulations for Safety Relevant Tasks, - Final Disposal: From Scientific Basis to Application, - Characteristics of a High Reliability Organization (HRO) Considering Experience Gained from Events at Nuclear Power Stations, and - Nuclear Competence in Germany and Europe have been covered in atw 7, 8/9, 10 and 11 (2011). (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Fung Kon Jin, P.H.P., E-mail: p.fungkonjin@amc.uva.nl [Trauma Unit Department of Surgery, Academic Medical Center, Amsterdam (Netherlands); Dijkgraaf, M.G.W., E-mail: m.dijkgraaf@amc.uva.nl [Department of Clinical Epidemiology, Biostatistics, and Bioinformatics, Academic Medical Center, Amsterdam (Netherlands); Alons, C.L., E-mail: clalons@few.vu.nl [Department of Mathematics, VU University Amsterdam, Amsterdam (Netherlands); Kuijk, C. van, E-mail: c.vankuijk@vumc.nl [Department of Radiology, VU Medical Center, Amsterdam (Netherlands); Beenen, L.F.M., E-mail: l.beenen@amc.uva.nl [Department of Radiology, Academic Medical Center, Amsterdam (Netherlands); Koole, G.M., E-mail: koole@few.vu.nl [Department of Mathematics, VU University Amsterdam, Amsterdam (Netherlands); Goslings, J.C., E-mail: j.c.goslings@amc.uva.nl [Trauma Unit Department of Surgery, Academic Medical Center, Amsterdam (Netherlands)

    2011-11-15

    Introduction: 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 an institutional perspective on process quality. Materials and methods: A total of six clinically relevant scenarios with variables that included the number of CT scanners, CT scanner location, and different patient categories (regular, urgent, and trauma patients) were evaluated using computer simulation. Each scenario was simulated using institutional data and was assessed for patient waiting times, idle time of CT scanners, and overtime due to scheduling. The best 2- and 3-scanner scenarios were additionally evaluated with the ATW-concept. Results: Based on institutional data, the best 2-scanner scenario distributes all 3 patient categories over both scanners and plans 4 urgent patients per hour while locating both scanners outside of the trauma room. The best 3-scanner scenario distributes urgent and regular patients over all 3 scanners and trauma patients on only 1 scanner and locates all CT scanners outside of the trauma room. The ATW concept reduces waiting times and overtime, while increasing idle time. Conclusion: Choosing the optimal planning and distribution strategies depends on the number and location of available CT scanners, along with number of trauma, urgent and regular patients. The Amsterdam Trauma Workflow concept could provide institutions with the ability of early CT scanning in trauma patients without influencing regular and urgent CT scanning.

  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. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  15. Transient tests on blower trip and rod removal at the HTR-10

    Energy Technology Data Exchange (ETDEWEB)

    Hu Shouyin [Institute of Nuclear Energy Technology, Tsinghua University, P.O. Box 1021, Beijing 102201 (China)]. E-mail: hyyhtr@tsinghua.edu.cn; Wang Ruipian [Institute of Nuclear Energy Technology, Tsinghua University, P.O. Box 1021, Beijing 102201 (China); Gao Zuying [Institute of Nuclear Energy Technology, Tsinghua University, P.O. Box 1021, Beijing 102201 (China)

    2006-03-15

    Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes. Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results. Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 deg. C during two tests. The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW. The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.

  16. 七、渔业生态环境

    Institute of Scientific and Technical Information of China (English)

    1994-01-01

    940356 硫丹在方鳍鱼体中的生物累积和排出=Bioaccumulation and elimination of endosul-fan in the fish yellow tetra(Hyphessobrycon bifasci-atws)[刊.英]/Jonsson C M//Bull.Environ.Con- tam.Toxic..—1993.50(4).—572~577硫丹农药是一种广泛应于可可、咖啡、棉花和大豆等农作物的杀虫剂,尽管作者已知硫丹对鱼类有较高的毒性,但硫丹在鱼体中

  17. A Numerical Study of the Plata River Plume Along the Southeastern South American Continental Shelf

    Science.gov (United States)

    2005-01-01

    banda de baixa salinidade formou-se desde o estu•.rio atW 30"N na plataforma continental sul brasileira. 0 efeito de ventos de sudoeste, que causam...distribuiqdo de baixa salinidade 6 muito mais sensivel A direggo dos ventos do que A descarga fluvial. Ventos de sudoeste sao capazes de advectar o sinai...de baixa salinidade ao longo da costa. JA os ventos de nordeste demonstram ser eficientes na erosgo da pluma, que 6 destacada da costa por deriva de

  18. Protozoa and their bacterial prey colonize sterile soil fast

    DEFF Research Database (Denmark)

    Altenburger, Andreas; Ekelund, Flemming; Jacobsen, Carsten Suhr

    2010-01-01

    We know little about the ability of protozoa to colonize soils, including their successional patterns. To elucidate this issue, we investigated in which order different protozoan morpho-types colonize sterile soil. We used sterilized soils with different carbon content, and exposed them to the at......We know little about the ability of protozoa to colonize soils, including their successional patterns. To elucidate this issue, we investigated in which order different protozoan morpho-types colonize sterile soil. We used sterilized soils with different carbon content, and exposed them...

  19. Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

    Science.gov (United States)

    Semidotskiy, I. I.; Kurskiy, A. S.

    2013-12-01

    The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.

  20. Special MAFIA postprocessors for the analysis of RF structures

    Science.gov (United States)

    Browman, M. J.

    1992-08-01

    This paper describes three stand-alone programs that use the electromagnetic fields generated by the MAFIA 2.04 codes to analyze radio-frequency (RF) cavities. Illustrations are provided that show how these codes are used to do the following: (1) analyze the effect of the coupling slots on the electric and magnetic fields of the linacs for the APLE Prototype Experiment (APEX) and the Advanced Free-Electron Laser (AFEL); (2) verify the Panofsky-Wenzel theorem for a high-energy deflecting cavity proposed for the Accelerator Transmutation of Waste (ATW) project; and (3) study the effectiveness of that deflecting cavity.

  1. 2009 Annual meeting on nuclear technology. Pt. 2. Topical sessions; Jahrestagung Kerntechnik 2009. T.2. Fachsitzungsberichte

    Energy Technology Data Exchange (ETDEWEB)

    Fazio, Concetta [Forschungszentrum Karlsruhe/KIT, Eggenstein-Leopoldshafen (Germany); Delpech, Marc [Centre CEA de Saclay (Essonne), Gif-sur-Yvette (France); Schaffrath, Andreas [TUeV NORD SysTec GmbH und Co. KG, Abt. Sicherheitsanalyse und Systemtechnik (ETB), Hamburg (Germany)

    2009-10-15

    Summary report on the Topical Sessions of the Annual Conference on Nuclear Technology held in Dresden, May 12 to 14, 2009: Flexible Concept for Sustainable Use of Nuclear Energy and Waste Minimization; Thermohydraulic Experiments for Reactors of the Second and Third Generation. The reports on the Topical Sessions: Advances in the Development of Integrated Management Systems for the Optimisation of Safety and Operational Availability of Nuclear Power Stations (Dr. Markus Nie and Dipl.-Ing. Karl Ramler), and; Fuel Elements: Zero Failure - Road to the Target (Dipl.-Ing. Andreas Huettmann) have been covered in atw 8/9 (2009). (orig.)

  2. 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)

  3. Innovative robotics and ultrasonic techniques; Innovative Robotik und Ultraschalltechnik. Pruefung von Reaktordruckbehaeltern in Siedewasser- und Druckwasserreaktoranlagen

    Energy Technology Data Exchange (ETDEWEB)

    Dirauf, F.; Fischer, E.; Gohlke, B. [Siemens AG, Erlangen (Germany). Abt. NW-P

    1998-06-01

    In-service inspections of the reactor pressure vessels of nuclear power plants not only impose very stringent requirements in terms of technology and quality, but increasingly must meet economic criteria as well. The use of US phased arrays opens up possibilities of minimizing the inspection system and thus reducing the number of inspection runs. Moreover, integration of US phased array technology into the Saphir system allows standard US equipment to be employed in inspecting complicated geometries (such as nozzles) which so far had been inspected only with specialized systems. Advanced inspection robots are under development for both BWR and PWR plants which are more flexible in use and more cost effective than conventional manipulators. (orig./GL) [Deutsch] Die Wiederkehrende Pruefung der Reaktordruckbehaelter von Kernkraftwerken stellt nicht nur hohe Anforderungen an Technologie und Qualitaet, sondern zunehmend ruecken auch oekonomische Kriterien in den Vordergrund. Der Einsatz der US-Gruppenstrahlertechnik eroeffnet die Moeglicheit der Pruefsystemminimierung und somit eine Reduzierung der Prueffahrten. Weiterhin schafft die Integration der US-Gruppenstrahlertechnik in das Saphir-System die Moeglichkeit, schwierige Geometrien (z.B. Stutzen), die bisher nur mit speziellen Systemen geprueft wurden, mit dem Standard-US-Geraet zu pruefen. Sowohl fuer SWR- als auch fuer DWR-Kernkraftwerke werden fortschrittliche Pruefroboter entwickelt, die flexibler einsetzbar, schneller und kostenguenstiger sind als herkoemmliche Manipulatoren. (orig./GL)

  4. 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.)

  5. Information policy of a Swiss nuclear power plant; Information der Medien und der Oeffentlichkeit durch den Anlagenbetreiber

    Energy Technology Data Exchange (ETDEWEB)

    Erne, L. [Kernkraftwerk Leibstadt (Switzerland)

    1997-12-31

    The Leibstadt NPP is situated on the river Rhine between Schaffhausen and Basel. It is the policy of the Leibstadt Plant Management to inform the public and the media in a timely and accurate manner during normal operation as well as in the case of unforeseen eventualities. We also inform the authorities of the communities in Southern Germany in the vicinity of the plant. Detailed procedures have been set up to provide a high standard of media service. The PR group is part of the emergency organisation and is supported by other technical groups. Modern communication technology, i.e. fax, teletext and internet are available. Our experience has shown that the early stages of an incident can be very hectic. The relevant information has to be collected internally and at the same time the media has to be presented with accurate communiques. (orig.) [Deutsch] Das Kernkraftwerk Leibstadt, gelegen auf der Schweizer Seite des Hochrheins, bekennt sich zu einer raschen, aktiven, transparenten und grenzueberschreitenden Informationspolitik sowohl im Normalbetrieb wie bei ungeplanten Ereignissen. Das Informationsverhalten ist festgelegt in detaillierten Vorschriften. Die Informationsstelle ist eingebunden in die Notfallorganisation mit Unterstuetzung von entsprechenden Dienstgruppen. In der Praxis nutzt das Werk moderne Kommunikationsmittel wie Telefax, Teletext oder neu Internet. Erfahrungen zeigen, dass in der Startphase eines Ereignisses der Zeitdruck enorm ist, waehrend intern gleichzeitig zuerst gesicherte Informationen zu beschaffen sind. (orig.)

  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. Beryllium detection in human lung tissue using electron probe X-ray microanalysis.

    Science.gov (United States)

    Butnor, Kelly J; Sporn, Thomas A; Ingram, Peter; Gunasegaram, Sue; Pinto, John F; Roggli, Victor L

    2003-11-01

    Chronic berylliosis is an uncommon disease that is caused by the inhalation of beryllium particles, dust, or fumes. The distinction between chronic berylliosis and sarcoidosis can be difficult both clinically and histologically, as both entities can have similar presentations and exhibit nonnecrotizing granulomatous inflammation of the lungs. The diagnosis of chronic berylliosis relies on a history of exposure to beryllium, roentgenographic evidence of diffuse nodular disease, and demonstration of beryllium hypersensitivity by ancillary studies, such as lymphocyte proliferation testing. Additional support may be gained by the demonstration of beryllium in lung tissue. Unlike other exogenous particulates, such as asbestos, detection of beryllium in human lung tissue is problematic. The low atomic number of beryllium usually makes it unsuitable for conventional microprobe analysis. We describe a case of chronic berylliosis in which beryllium was detected in lung tissue using atmospheric thin-window energy-dispersive X-ray analysis (ATW EDXA). A woman with a history of occupational exposure to beryllium at a nuclear weapons testing facility presented with progressive cough and dyspnea and a nodular pattern on chest roentgenograph. Open lung biopsy showed nonnecrotizing granulomatous inflammation that was histologically indistinguishable from sarcoidosis. Scanning electron microscopy and ATW EDXA demonstrated particulates containing beryllium within the granulomas. This application of EDXA offers significant advantages over existing methods of beryllium detection in that it is nondestructive, more widely available, and can be performed using routine paraffin sections.

  8. Cube-textured metal substrates for reel-to-reel processing of coated conductors

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, A.C.

    2012-09-15

    This thesis presents the results of a study aimed at investigating important fabrication aspects of reel-to-reel processing of metal substrates for coated conductors and identifying a new substrate candidate material with improved magnetic properties. The effect of mechanical polishing on surface roughness and texture in Ni-5at.%W tapes in the cold-rolled condition was studied as a function of polishing grade. The surface roughness of the tape in the polished and annealed condition, and after subsequent coating with a Gd{sub 2}Zr{sub 2}O{sub 7} buffer layer was investigated taking grain boundaries into account. It was observed that the initial mean surface roughness decreased after annealing except after very fine polishing. Additionally, the roughness of the buffer layers were found to increase slightly for the fine polished substrates. Grain boundary grooving was observed to impose a lower limit for the mean surface roughness. Fractions of cube texture within deviations of 5 deg. from the ideal cube orientation, in the annealed substrates, were found to be very sensitive to the surface roughness before annealing. Microstructure, texture and topography were studied in a strongly cube-textured Ni-5at.%W substrate before and after an additional annealing (condition A1 and A2, respectively) simulating a buffer layer crystallisation heat treatment. Condition A1 was characterised by a high fraction of cube texture, a high fraction of low angle grain boundaries and a low fraction of {Sigma}3 boundaries. A strong correlation was observed between the grain boundary groove depth and boundary type. Coherent twin boundaries and low angle grain boundaries were characterised by the smallest average groove depth while significantly deeper grooves were observed at other boundary types. A similar correlation was observed between the inclination angle at groove walls and the boundary type. The microstructure was slightly coarser in condition A2 and it was accompanied by a cube

  9. The Clinical Use of Left Ventricular Twist Degree in Chronic Heart Failure Subjects by Three-dimensional Ultrasound Speckle Tracking Imaging%三维斑点追踪技术在慢性心力衰竭患者左室扭转运动中的应用研究

    Institute of Scientific and Technical Information of China (English)

    张艳丽; 王小丛; 赵丽荣; 装莉平; 于微

    2012-01-01

    Objective This study was performed to assess left ventricle twist degree in patients with chronic heart failure by three-dimensional ultrasound speckle tracking imaging. Methods The apical 4-chamber and 2-chamber views were acquired in thirty-two patients with chronic heart failure and thirty-three healthy volunteers .using 3D-trace software to measure values of left ventricle end-diastolic volumes (LVEDV) , end-systolic volumes(LVESV) ,left ventricular ejection fraction (LVEF),basal segment twist degree(BTW),middle segment twist degree(MTW) .apical segment twist degree( ATW) ,left ventricular global twist degree(LVTW). Values were compared in two groups, the correlations between LVEF and LVTW,BTW,MTW, ATW were analyzed respectively. Results LVEF,LVTW,MTW, ATW in CHF patients were lower than the control group .the correlations between BTW,MTW, ATW,LVTW and LVEF were found (0. 557,0. 926,0. 932,0. 945. P<0. 01 for all). Conclusions The left ventricular function was impaired in patients with CHF. The left ventricular twist can be studied by three-dimensional ultrasound speckle tracking imaging, which would be a new tool for the evaluation of left ventricular systolic function.%目的 应用三维斑点追踪显像技术研究慢性心力衰竭患者(CHF)左室扭转的运动特征,探讨其临床价值.方法 CHF组患者30例,年龄匹配的健康志愿者(对照组)33例,采集标准的四腔心及两腔心切面,进行全容积图像存储,应用3D-trace软件进行脱机分析,软件自动分析计算左心室舒张末期容积(LVEDV),左室收缩末期容积(LVESV),左室射血分数(LVEF),左室基底段收缩期扭转角度峰值(BTW),中间段收缩期扭转角度峰值(MTW),心尖段收缩期扭转角度峰值(ATW),左室整体收缩期扭转角度峰值(LVTW).结果 CHF组LVEF,LVTW,MTW,ATW均较正常组减低,BTW、MTW、ATW、LVTW与LVEF之间有明显的相关性,相关系数分别为0.557,0.926,0.932,0.945.结论 CHF患者左心收缩功能明显降

  10. 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

  11. Planning and reporting of Russian transmutation research projects within ISTC. Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    Conde, H. [Uppsala Univ. (Sweden). Dept. of Neutron Research; Gudowski, W. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Reactor and Neutron Physics; Liljenzin, J.O. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry; Mileikovsky, C. [Pully (Switzerland)

    1998-11-01

    The present report about phase 2 of the SKI project on Planning and Reporting of Russian Transmutation Research Projects within ISTC is an update of the information given in the SKI report no 97:15 (Feb 1997) about phase 1 of the same project. The background information is partly repeated in the present report to avoid that the reader has to go back to the report of Phase 1 for information about the basis for the project. USA, EU, Japan, Republic of Korea and Norway are at present supporting the International Scientific and Technical Center (ISTC) in Moscow. The Centre gives funds to research projects of civilian interest to former nuclear weapon laboratories to counteract the risk of nuclear weapon proliferation by the emigration of former USSR technical and scientific experts to `border countries` which are aiming towards the development of nuclear weapons. Before Sweden and Finland entered the EU, both countries gave national support to ISTC, in the case of Sweden 4 MUSD. Some of the projects which were funded by the Swedish national support to ISTC are still in progress. Nuclear technical concepts (i.e. Accelerator Transmutation of Nuclear Waste, ATW) have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The named Russian experts are knowledgeable and well equipped of doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been initiated, and further ones have been proposed, to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. A similar centre STCU (The Scientific and Technical Centre of the Ukraine) has been set up in Kiev. Sweden has been active in promoting this Centre, which is supported by USA, Japan, Canada and recently also by EU. The present report describes the

  12. 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

  13. 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.)

  14. Inhibition of interleukin-13 gene expression in T cells through GATA-3 pathway by arsenic trioxide

    Institute of Scientific and Technical Information of China (English)

    YAO Xin; HE Hai-yan; YANG Yan; DAI Shan-lin; SUN Pei-li; YIN Kai-sheng; HUANG Mao

    2008-01-01

    @@ Arsenic trioxide (AT) has a long history of use in both traditional Chinese medicine and in modern medicine in asthma therapy.Recently,Yin et al1 found that AT even at small doses reduced the airway inflammation of sensitized guinea pigs.However the mechanism underlying this is still largely unknown.Interleukin 13 (IL-13),as one of the important TH2 cytokines,plays an important role in asthma pathogenesis through promoting eosinophilic inflammation,mucus secretion and airway hyperresponsiveness.2 To further explore the molecular anti-inflammatory basis of AT,we employed Hut-78 cells,a human T cell line,with activation via CD3/CD28 receptors to mimick in vivo co-stimulation to investigate the effect of AT on IL-13 transcription.

  15. 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.

  16. Comparative study of different models of transportation of boron in the codes Thermohydraulic TRAC-BF1, TRACE and RELAP; Estudio Comparativo de Diferentes Modelos de Transporte de Boro en los Codigos Termohidraulicos TRAC-BF1, TRACE y RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Jambrina, A.; Solar, A.; Barrachina, T.; Miro, R.; Verdu, G.; Concejal, A.

    2013-07-01

    In BWR 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. The boron transport model implemented in TRAC-BF1 code is based on a first order accurate upwind difference scheme. Four numerical schemes that solve the boron transport model have been analyzed and compared with the analytical solution that provides the Burgers equation: first order Upwind, second order Godunov, second-order modified Godunov and a third-order QUICKEST using the ULTIMATE universal limiter. The modified Godunov scheme has been implemented in TRAC-BF1 source code. The results using these new schemes are presented in this paper.

  17. THERMAL HYDRAULIC ANALYSIS OF A LIQUID-METAL-COOLED NEUTRON SPALLATION TARGET

    Energy Technology Data Exchange (ETDEWEB)

    W. GREGORY; R. MARTIN; T. VALACHOVIC

    2000-07-01

    We have carried out numerical simulations of the thermal hydraulic behavior of a neutron spallation target where liquid metal lead-bismuth serves as both coolant and as a neutron spallation source. The target is one of three designs provided by the Institute of Physics and Power Engineering (IPPE) in Russia. This type of target is proposed for Accelerator-driven Transmutation of Waste (ATW) to eliminate plutonium from hazardous fission products. The thermal hydraulic behavior was simulated by use of a commercial CFD computer code called CFX. Maximum temperatures in the diaphragm window and in the liquid lead were determined. In addition the total pressure drop through the target was predicted. The results of the CFX analysis were close to those results predicted by IPPE in their preliminary analysis.

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Choi, J. T.; Kim, I. J. [Sunmoon Univ., Asan (Korea, Republic of); Lee, S. H.; Baek, W. P.; Yang, S. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

    1998-06-15

    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 as follows : reliability of risk significant SSCs established during design stage must be maintained through the operating life of the plant, currently used classification method of plant conditions and safety requirement were reviewed, and a quantitative classification method is needed to be developed further, the basic regulatory directions are proposed for multiple failures such as SBO, TLOFW, multiple SGTR and ATWS, safety requirements are proposed for survivability/availability of severe accident mitigation design features by 5 items if basic requirements, selection of initial event, identification of available equipment and instruments, identification of environmental conditions and verification methods.

  20. 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...... film exhibit high homogeneity in terms of biaxial texture over the tape. The average values of full width at half maximum of in-plane and out-of-plane texture are 7.2° and 6.1° in Ni5W substrate, 7.6° and 6.1° in CeO2 buffer layer, respectively, all of those with a small standard deviation...

  1. 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.

  2. LOFA and RIA analysis of the Indonesian Multipurpose research reactor RSG-GAS (1)

    Energy Technology Data Exchange (ETDEWEB)

    Endiah Puji Hastuti; Hudi Hastowo; Iman Kuntoro [Center for Multipurpose Reactor, National Atomic Energy Agency (BATAN), Puspiptek, Serpong, Tangerang (Indonesia)

    1999-07-01

    Investigation on accident of the Indonesian Multipurpose research reactor RSG-GAS has been performed by computer simulation technique. Two groups of transients were considered, namely transient due to loss of primary cooling system (LOFA) and power excursion due to reactivity insertion (RIA). In such a transient condition, the Common Mode Failure (CMF) is considered and it will induce a situation so called unprotected transient or Anticipated Transient Without Scram (ATWS). RELAP5, PARET-ANL and EUREKA-2RR computer packages have been applied for these analyses. Simulations result done using these computer packages showed that in the occurrence of LOFA and RIA, failure on fuel elements is limited to the region with the highest power factor. (author)

  3. 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)

  4. Flight Research and Validation Formerly Experimental Capabilities Supersonic Project

    Science.gov (United States)

    Banks, Daniel

    2009-01-01

    This slide presentation reviews the work of the Experimental Capabilities Supersonic project, that is being reorganized into Flight Research and Validation. The work of Experimental Capabilities Project in FY '09 is reviewed, and the specific centers that is assigned to do the work is given. The portfolio of the newly formed Flight Research and Validation (FRV) group is also reviewed. The various projects for FY '10 for the FRV are detailed. These projects include: Eagle Probe, Channeled Centerbody Inlet Experiment (CCIE), Supersonic Boundary layer Transition test (SBLT), Aero-elastic Test Wing-2 (ATW-2), G-V External Vision Systems (G5 XVS), Air-to-Air Schlieren (A2A), In Flight Background Oriented Schlieren (BOS), Dynamic Inertia Measurement Technique (DIM), and Advanced In-Flight IR Thermography (AIR-T).

  5. 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.)

  6. TomoWELD. Precise detection of weld defects; TomoWELD. Defekte in Schweissnaehten praezise erkennen

    Energy Technology Data Exchange (ETDEWEB)

    Walter, David [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2016-06-15

    Nuclear power plants are complex and technically elaborate systems whose aim is to produce electricity. They must meet the highest safety requirements. Within the reactors, nuclear reactions and radioactive transformations release energy which is used to evaporate water. The steam generated drives turbines that in turn are coupled with generators which convert the kinetic energy provided by the turbines into electrical energy. The process is easy to illustrate but difficult to control and requires technical equipment such as kilometre-long pipe systems. Austenitic steel is frequently used for this purpose because of its high strength and corrosion resistance. The individual pipe components are joined by welding. However, welds may contain hidden defects. Cracks, lack of fusion or pore nests that can remain undetected may have catastrophic consequences. Therefore, all welds in a nuclear power plant, without exception, must be checked. Approved non-destructive methods use ultrasound and X-ray. The technology developed at BAM is called TomoWELD. [German] Kernkraftwerke sind komplexe und technisch aufwendige Anlagen zur Gewinnung von Elektrizitaet. Sie muessen allerhoechsten Sicherheitsanspruechen genuegen. Die bei Kernreaktionen und radioaktiven Umwandlungen freiwerdende Energie wird genutzt, um Wasser zu verdampfen. Der Dampf treibt Turbinen an und die wiederum sind mit Generatoren gekoppelt, welche die durch die Turbinen bereitgestellte kinetische Energie in elektrische Energie umwandeln. Der Prozess laesst sich einfach darstellen, ihn zu steuern ist allerdings kompliziert und erfordert weitere technische Komponenten, wie beispielsweise kilometerlange Rohrleitungssysteme. Wegen seiner hohen Festigkeit sowie Korrosionsbestaendigkeit wird oft austenitischer Stahl dafuer verwendet. Gefuegt werden die einzelnen Rohrteile durch Schweissen. Doch Schweissnaehte koennen viele verborgene Defekte enthalten. Bleiben Risse, Bindefehler oder Porennester unentdeckt, kann das

  7. Cooperation in the nuclear field among the states of Eastern Europe; Die kerntechnische Zusammenarbeit der Staaten Osteuropas

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, H. [Tetra Energie Technologie Transfer GmbH, Berlin (Germany); Kraemer, J. [Tetra Energie Technologie Transfer GmbH, Berlin (Germany)

    1996-11-01

    Since the re-orientation of the societies of Central and Eastern Europe on the basis of a free and democratic order, the West has launched a number of support programs for safe operation of the nuclear power plants run in these countries. It should not be overlooked, however, that a traditionally close cooperation in all areas of the military and peaceful uses of nuclear power had existed among the Socialist countries, mostly dominated by the Soviet Union, i.e. Russia. After the initiation of the process of democratization, the turnaround to the West was overemphasized initially, but in the meantime the old ties of cooperation have been strengthened again, though with more sovereignty of the participating nations. Russia is striving to maintain her hegemony and exploit the findings of recent nuclear research in order to boost her balance of trade by exports. (orig.) [Deutsch] Seit sich die Gesellschaften der Staaten Mittel- und Osteuropas neu orientieren, und das auf der Basis einer freiheitlich-demokratischen Grundordnung, wurden im Westen unterschiedlichste Unterstuetzungsprogramme fuer einen sicheren Betrieb der in diesen Laendern betriebenen Kernkraftwerke aufgelegt. Dabei darf nicht ausser acht gelassen werden, dass traditionell eine enge Zusammenarbeit auf allen Gebieten der militaerischen und friedlichen Nutzung der Kernenergie in den sozialistischen Staaten existierte, zumeist unter Dominanz der Sowjetunion, und hier Russlands. Nach Einleitung des Demokratisierungsprozesses wurde die Hinwendung zum Westen zunaechst ueberbetont; inzwischen aber wurden die gewachsenen Kooperationen wieder gefestigt, allerdings unter groesserer Souveraenitaet der beteiligten Nationen. Russland ist bestrebt, seine Hegemonie zu behaupten und die Ergebnisse neuerer kerntechnischer Forschung zu nutzen, um durch Exporte seine Handelsbilanz zu verbessern. (orig.)

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

    Directory of Open Access Journals (Sweden)

    Majid Motaghinejad

    2015-05-01

    Full Text Available 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 (P<0.05 for groups received diazepam and treated by forced exercise and P<0.001 for group under treatment diazepam + 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.

  9. 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.

  10. Analyses of Design Extended Condition Events for the Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Jeong, Taekyung; Lee, Kwilim; Jeong, Jaeho; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). For example, the CRSS in the PRISM designed based on the 0.4 $ reactivity insertion, which is analyzed with safety analysis of UTOP event. Moreover, the sensitivity tests for weighting factor in the core radial expansion reactivity feedback model were also carried out for all ATWS events. Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W{sub GP}, which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W{sub GP} is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate expansion in the ULOHS is predominant.

  11. 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.

  12. Safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-04-01

    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, 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 safety analysis are described in chapter 2. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code 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 performance analysis results of the KALIMER containment dome are described along with the HCDA accident scenario and source terms. The major containment parameters of peak pressure and peak temperature have been calculated using the CONTAIN-LMR code. Radiological consequence has been evaluated by the MACCS code. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using SCHAMBETA code developed in the framework of the modified bethe-tait method. Work energy potentials based arising from the sodium expansion as well as the isentropic fuel expansion are then calculated to evaluate the structural integrity of the reactor vessel, reactor internals and primary coolant system of KALIMER.

  13. Preliminary safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-06-01

    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. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER 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 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 to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are 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 energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  14. 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.

  15. 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.

  16. Furrow-irrigated chufa crops in Valencia (Spain. I: Productive response to two irrigation strategies

    Directory of Open Access Journals (Sweden)

    N. Pascual-Seva

    2013-01-01

    Full Text Available Chufa (Cyperus esculentus L. var. sativus Boeck. is an important vegetable crop in Valencia (Spain, where its tubers are used to produce a refreshing drink called 'horchata'. Water is relatively inexpensive, there are no data regarding the volumes of water used to grow chufa, and the irrigation water use efficiency (IWUE has neither been determined. The aim of this research was to compare the productive responses of the chufa crop to two irrigation strategies (IS. The volumetric soil water content (VSWC was monitored with capacitance sensors. Trends in VSWC were used to determine the in situ field capacity (FC, beginning each irrigation event when the VSWC reached either approximately 45% (H1 or 60% (H2 of the FC at a soil depth of 0.10 m. The experiments were conducted over three consecutive seasons. An area velocity flow module measured the water flow. The yields, the water volumes used, and the IWUE were calculated. Plants were periodically sampled and the harvest index and relative growth rate were determined. The yield was affected by the year and by the IS. The greatest yields were obtained with the H2 strategy (on average 2.18 kg m-2 for H2 vs. 1.94 kg m-2 for H1; p≤0.01, and the average tuber weight (ATW was affected (p≤0.01 by the year and IS interaction. IWUE was affected by the year, and none of the considered factors affected the harvest index (p≤0.05. It can be concluded that maintaining a higher VSWC would increase both yield and ATW without affecting IWUE.

  17. 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.

  18. Simulator training. Requirements and ways to meet them; Die Simulatorschulung. Anforderungen und deren Realisierung

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, E. [Gesellschaft fuer Simulatorschulung mbH (French Guiana), Essen (Germany) KSG Kraftwerks-Simulator-Gesellschaft mbH, Essen (Germany)

    1997-02-01

    The excellent operating performance of German nuclear power plants year by year attests to the high reliability and safety of these plants. On the one hand, this is due to the mature technical plant concept and, on the other hand, to the safety culture of plant operation, i.e. the qualified and careful work of the plant personnel running and monitoring the plants and keeping them in top condition technically. Special responsibility rests on the shift personnel operating plants around the clock. It is for this reason that operators take very seriously the efforts by these staff members to acquire and preserve a high level of training. Training by simulators, as organized centrally in Essen for twenty-two nuclear generating units in Germany, the Netherlands, and Switzerland, plays a major role in these activities. The special characteristic of simulator training is the combination of two things, namely technology and human behavior. The power plant processes, with all their complexities, must be conveyed to human operators, with all their skills traits and weaknesses, by means of appropriate simulation. The requirements resulting from this need, and the solutions adopted at the Essen Simulator Center, are described in this article. (orig.) [Deutsch] Die deutschen Kernkraftwerke beweisen mit ihren hervorragenden Betriebsergebnissen jedes Jahr erneut ihre hohe Zuverlaessigkeit und Sicherheit. Dies beruht zum einen auf dem technisch ausgereiften Anlagenkonzept und zum anderen auf der Sicherheitskultur des Betriebes, d.h. der qualifizierten und sorgfaeltigen Arbeit des Betriebspersonals, welches das Kraftwerk betreibt, ueberwacht und in technisch einwandfreiem Zustand erhaelt. Eine besondere Verantwortung hat das Personal der Schichten, welche die Anlage rund um die Uhr `fahren`. Enstprechend ernst nehmen deshalb die Betreiber den Erwerb und den dauerhaften Erhalt eines hohen Ausbildungsstandes dieser Mitarbeiter. Die Schulung an Simulatoren - wie sie zentral in Essen

  19. Nuclear power plant personnel errors in decision-making as an object of probabilistic risk assessment. Methodological extensions on the basis of a differentiated analysis of safety-relevant goals; Entscheidungsfehler des Betriebspersonals von Kernkraftwerken als Objekt probabilistischer Risikoanalysen; Methodische Erweiterungen auf der Basis einer differenzierten Betrachtungsweise sicherheitsgerichteter Ziele

    Energy Technology Data Exchange (ETDEWEB)

    Reer, B.

    1993-09-01

    wird eine neuartige Fragetechnik entwickelt und gezeigt, dass solche Fehler als Rueckwirkungen untergeordneter Ziele auftreten koennen. Solche Rueckwirkungen sind ueber eine differenzierte Betrachtung sicherheitsgerichteter Ziele identifizierbar. Zur Quantifizierung wird eine neue Methode entwickelt, mit der sich situationsspezifisch Wahrscheinlichkeiten fuer Entscheidungsfehler schaetzen lassen. Es gelingt, die Faktoren Konflikt und Aehnlichkeit so zu operationalisieren, dass sie mit den Informationen, die einem PRA-Anwender ueblicherweise zur Verfuegung stehen, quantitativ zugaenglich sind. Das Quantifizierungs verfahren basiert auf Extra- und Interpolationen zu den wenigen Daten, die zur Zeit ueber Entscheidungsfehler von Operateuren existieren. Ausserdem wird fuer passive Entscheidungsfehler (Unterlassungen notwendiger Handlungen) ein neuartiger Modellansatz vorgestellt, der solche Fehler ueber Verzoegerungszeiten quantifiziert. Die praktische Durchfuehrbarkeit dieses dynamischen Ansatzes wird am Beispiel einer probabilistischen Analyse der konkreten Operateurmassnahmen gezeigt, die waehrend des Stoerfalls im Kernkraftwerk Davis-Besse (1985) angefordert wurden. Die Erweiterungen der klassischen MMSA-Methodik werden am Beispiel der Nachwaermeabfuhr (NWA) des HTR-500 angewendet. Entscheidungsfehler (als Ursachen ungeplanter Handlungen) werden systematisch und umfassend beruecksichtigt. Fuenf zusaetzliche Entscheidungsfehler werden identifiziert.

  20. Does electricity from nuclear power stand a chance in competition?; Hat Kernenergie-Strom eine Chance im Wettbewerb?

    Energy Technology Data Exchange (ETDEWEB)

    Hohlefelder, W. [PreussenElektra AG, Hannover (Germany); PreussenElektra Netz GmbH und Co. KG, Hannover (Germany)

    2000-03-01

    , ausstiegsorientierten Vollzug oder durch kuenstliche wirtschaftliche Zusatzbelastungen. 3. Der Kostenvorsprung zu anderen Erzeugungstechniken verringert sich. Deshalb ist ein konsequentes Kostenmanagement unerlaesslich, aber nur soweit die Sicherheit der Anlage unangetastet bleibt. 4. Bei einem Ausstieg wuerde die 'deutsche' Kernenergie jedenfalls zum grossen Teil durch auslaendische ersetzt. Dies erhoeht den Anreiz, jene so lange wie moeglich zu betreiben. 5. In voellig liberalisierten Maerkten ist es aus wirtschaftlichen Gruenden schwierig, neue Kernkraftwerke zu errichten. Es findet eine einseitige Ausrichtung auf einen, naemlich den kostenguenstigsten und am wenigsten kapitalintensiven Energietraeger statt. Darin liegt ein erhebliches Versorgungsrisiko. Unbeschadet der grundsaetzlichen Entscheidung fuer die Liberalisierung des Strommarktes ist daher eine Korrektur der politisch gesetzten Rahmenbedingungen, um Fehlentwicklungen zu vermeiden, auf mittlere Sicht zu erwarten. (orig.)

  1. Sensitivity Tests for the Unprotected Events of the Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Lee, Kwilim; Jeong, Jaeho; Yu, Jin; An, Sangjun; Lee, Seung Won; Chang, Wonpyo; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Unprotected Transient Over Power, (UTOP), Unprotected Loss Of Flow (ULOF), and Unprotected Loss Of Heat Sink (ULOHS) are selected as ATWS events. Among these accidents, the ULOF event shows the lowest clad temperature. However, the ULOHS event showed the highest peak clad temperature, due to the positive CRDL/RV expansion reactivity feedback and insufficient DHRS capacity. In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W{sub GP}, which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W{sub GP} is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate

  2. New Fuel Pin Axial Expansion Reactivity Feedback Model in MARS-LMR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). For example, the CRSS in the PRISM designed based on the 0.4 $ reactivity insertion, which is analyzed with safety analysis of UTOP event. Moreover, the sensitivity tests for weighting factor in the core radial expansion reactivity feedback model were also carried out for all ATWS events. Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W{sub GP}, which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W{sub GP} is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate expansion in the ULOHS is predominant. These analysis results will give better

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

    Science.gov (United States)

    Pak, Chan-gi; Lung, Shu

    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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Successful retrieval of an unexpanded coronary stent from the left main coronary artery during primary percutaneous coronary intervention

    Directory of Open Access Journals (Sweden)

    Šalinger-Martinović Sonja

    2011-01-01

    Full Text Available Introduction. Dislodgement and embolization of the new generation of coronary stents before their deployment are rare but could constitute a very serious complication. Case Outline. We report a case of a stent dislodgement into the left main coronary artery during the primary coronary intervention of infarct related left circumflex artery in a patient with acute myocardial infarction. The dislodged and unexpanded bare-metal stent FlexMaster 3.0x19 mm (Abbot Vascular was stranded and bended in the left main coronary artery (LMCA, probably by the tip of the guiding catheter, but stayed over the guidewire. It was successfully retrieved using a low-profile Ryujin 1.25x15 balloon catheter (Terumo that was passed through the stent, inflated and then pulled back into the guiding catheter. After that, the whole system was withdrawn through the 6 F arterial sheath via the transfemoral approach. After repeated cannulation via the 6F arterial sheath, additional BMW and ATW guidewires were introduced into the posterolateral and obtuse marginal branches and a bare-metal stent Driver (Medtronic Cardiovascular Inc 3.0x18 mm was implanted in the target lesion. Conclusion. Stent dislodgement is a rare but potentially life-threatening complication of the percutaneous coronary intervention. This incident occurring in the LMCA in particular during an acute myocardial infarction requires to be urgently resolved. The avoidance of rough manipulation with the guiding catheter and delivery system may help in preventing this kind of complications.

  9. Preliminary Evaluation of the Diverse Protection System in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Taekyeong; Chang, Won-Pyo; Seong, Seung Hwan; Ahn, Sang June; Kang, Seok Hun; Choi, Chiwoong; Yoo, Jin; Lee, Kwi Lim; Lee, Seung Won; Jeong, Jae-Ho; Ha, Kwi-Seok [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The anticipated transient without scram (ATWS) is defined as an abnormal transient with failure of scram actuation. It is one of the “worst case” accident based on the United States Nuclear Regulatory Commission (U.S.NRC). Consideration frequently motivates the NRC to take regulatory action. An evaluation of this event is also a general requirement due to a potential safety issue that may lead to core damage under postulated condition. This paper estimated the set-points sensitivity test of the diverse protection system (DPS) related with unprotected events of the prototype generation-IV sodium cooled fast reactor (PGSFR) including unprotected transient over power (UTOP) and unprotected loss of flow (ULOF) by MARS-LMR code. The variation of the power to flow (P/Q) of UTOP and ULOF is illustrated to conduct the set-points sensitivity test of DPS. Also we estimated the effect of the DPS introduction after selecting UTOP, ULOF event as the unprotected events which are predicted to aggravate the events. This paper estimated the set-points sensitivity test of DPS related with unprotected events of PGSFR including UTOP and ULOF by MARS-LMR code. The results indicated that there is no significant difference in both RPS and DPS about the initiating time of each event. Therefore, this study found that the urgent manage for safety of the reactor when RPS failed is possible by the applying DPS.

  10. 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.

  11. Numerical Model for the Analysis of Coolability of a Particulate Debris Bed with Single Phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Cho, C. H.; Jeong, H. Y.; Chang, W. P.; Kwon, Y. M.; Lee, Y. B

    2008-01-15

    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. Accordingly, Numerical model development for the Analysis of coolability of a particulate debris bed with single phase flow was carried out for in-vessel retention of the core debris.

  12. Revaluation of a Coolability of a Packed Debris Bed with a Single Phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chungho; Suk, S. D.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B

    2008-09-15

    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. Accordingly, evaluation of coolability of a packed debris bed with single phase flow was carried out for proof of the in-vessel retention of the core debris.

  13. 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.

  14. Experimental Validation of RELAP5 and TRACE5 for Licensing Studies of the Boron Injection System of Atucha II

    Directory of Open Access Journals (Sweden)

    Alejandro I. Lazarte

    2011-01-01

    Full Text Available This paper presents an experimental validation of RELAP5 and TRACE5 for licensing studies of the Atucha II-PHWR nuclear power plant. A scaled experimental facility, representing the boron injection system of Atucha II, was built. The system has a fundamental importance for loss of coolant accidents (LOCA and anticipated transients without scram (ATWS. The experiment consists of the discharge of a tank that represents the boron tank filled with air or a mixture of air-water onto a discharge tank that represents the moderator tank. Both tanks are connected by a pipe which includes a valve and an orifice plate to model the pressure losses due to the fittings in the real system. The pressure and water level measured in the tanks are compared with the RELAP5 and TRACE5 predictions. The codes predict the pressure in the tanks accurately. However, both codes overpredict the heat transfer in the boron tank air-water interface which produces a greater expansion of the air which leads to a small discrepancy in the boron tank level prediction.

  15. Selection of flowing liquid lead target structural materials for accelerator driven transmutation applications

    Science.gov (United States)

    Park, John J.; Buksa, John J.

    1995-09-01

    The beam entry window and container for a liquid lead spallation target will be exposed to high fluxes of protons and neutrons that are both higher in magnitude and energy than have been experienced in proton accelerators and fission reactors, as well as in a corrosive environment. The structural material of the target should have a good compatibility with liquid lead, a sufficient mechanical strength at elevated temperatures, a good performance under an intense irradiation environment, and a low neutron absorption cross section; these factors have been used to rank the applicability of a wide range of materials for structural containment. Nb-1Zr has been selected for use as the structural container for the LANL ABC/ATW molten lead target. Corrosion and mass transfer behavior for various candidate structural materials in liquid lead are reviewed, together with the beneficial effects of inhibitors and various coatings to protect substrate against liquid lead corrosion. Mechanical properties of some candidate materials at elevated temperatures and the property changes resulting from 800 MeV proton irradiation are also reviewed.

  16. Updating the Finite Element Model of the Aerostructures Test Wing using Ground Vibration Test Data

    Science.gov (United States)

    Lung, Shun-fat; Pak, Chan-gi

    2009-01-01

    Improved and/or accelerated decision making is a crucial step during flutter certification processes. Unfortunately, most finite element structural dynamics models have uncertainties associated with model validity. Tuning the finite element model using measured data to minimize the model uncertainties is a challenging task in the area of structural dynamics. The model tuning process requires not only satisfactory correlations between analytical and experimental results, but also the retention of the mass and stiffness properties of the structures. Minimizing the difference between analytical and experimental results is a type of optimization problem. By utilizing the multidisciplinary design, analysis, and optimization (MDAO) tool in order to optimize the objective function and constraints; the mass properties, the natural frequencies, and the mode shapes can be matched to the target data to retain the mass matrix orthogonality. This approach has been applied to minimize the model uncertainties for the structural dynamics model of the Aerostructures Test Wing (ATW), which was designed and tested at the National Aeronautics and Space Administration (NASA) Dryden Flight Research Center (DFRC) (Edwards, California). This study has shown that natural frequencies and corresponding mode shapes from the updated finite element model have excellent agreement with corresponding measured data.

  17. A simple MOD method to grow a single buffer layer of Ce{sub 0.8}Gd{sub 0.2}O{sub 1.9} (CGO) for coated conductors

    Energy Technology Data Exchange (ETDEWEB)

    Liu Min, E-mail: Lm@bjut.edu.c [Institute for Superconducting and Electronic Materials, University of Wollongong, NSW 2522 (Australia); Key Laboratory of Advanced Functional Materials, Ministry of Education, College of Materials Science and Engineering, Beijing University of Technology, Beijing 100022 (China); Shi Dongqi, E-mail: dongqi@uow.edu.a [Institute for Superconducting and Electronic Materials, University of Wollongong, NSW 2522 (Australia); Suo Hongli; Ye Shuai; Zhao Yue; Zhu Yonghua [Key Laboratory of Advanced Functional Materials, Ministry of Education, College of Materials Science and Engineering, Beijing University of Technology, Beijing 100022 (China); Li Qi; Wang Lin; Jihyun Ahn [Institute for Superconducting and Electronic Materials, University of Wollongong, NSW 2522 (Australia); Zhou Meiling [Key Laboratory of Advanced Functional Materials, Ministry of Education, College of Materials Science and Engineering, Beijing University of Technology, Beijing 100022 (China)

    2009-03-15

    A single Ce{sub 0.8}Gd{sub 0.2}O{sub 1.9} (CGO) buffer layer was successfully grown on the home-made textured Ni-5 at.%W (Ni-5W) substrates for YBCO coated conductors by a simple metal-organic deposition (MOD) technique. The precursor solution was prepared using a newly developed process and only contained common metal-organic salts of both Ce and Gd dissolved into a propionic acid solvent. The precursor solution at 0.4 M concentration was spin coated on short samples of Ni-5W substrates and heat-treated at 1100 deg. C in a mixture gas of 5% H{sub 2} in Ar for an hour. X-ray studies indicated that the CGO films had good out-of-plane and in-plane textures with full-width-half-maximum values of 4.18 deg. and 6.19 deg., respectively. Atomic force microscope (AFM) investigations of the CGO films revealed that most of the grain boundary grooves on the Ni-5W surface were found to be well covered by CGO layers, which had a fairly dense and smooth microstructure without cracks and porosity. These results indicate that our MOD technique is very promising for further development of single buffer layer architecture for YBCO coated conductors, due to its low cost and simple process.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  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. Environmental effects on fatigue of steels for structural parts in water-steam-circuits of light water reactors. Considerations concerning the question of transferability of results from laboratory tests to real operating conditions; Der Einfluss des Mediums auf Ermuedungsvorgaenge in Staehlen fuer Strukturbauteile in Wasser-Dampf-Kreislaeufen von Leichtwasserreaktoren. Ueberlegungen zur Frage der Uebertragbarkeit von Ergebnissen aus Laborversuchen auf den realen Anlagenbetrieb

    Energy Technology Data Exchange (ETDEWEB)

    Roth, Armin [AREVA NP GmbH, Erlangen (Germany)

    2008-07-01

    .B. von Staehlen, beeinflussen kann. Schon vor einigen Jahrzehnten wurde experimentell nachgewiesen, dass z.B. Hochtemperaturwasser in Laborversuchen zur Untersuchung des Ermuedungsverhaltens von Staehlen zu deutlichen Effekten fuehrt. Dabei wird je nach Versuchsfuehrung entweder die Zeit bis zur Initiierung von Anrissen verkuerzt oder die Wachstumsgeschwindigkeit vorhandener Risse erhoeht. Dieser zu erwartende Einfluss des Mediums auf den Ermuedungsprozess wurde in den Anfaengen der Regulierung von Konstruktion und Auslegung von Bauteilen und Komponenten fuer Kernkraftwerke in den relevanten Regelwerken (z.B. ASME Boiler and Pressure Vessel Code, Section III) weltweit nicht explizit beruecksichtigt. Pauschal beruecksichtigt werden Umgebungseffekte dagegen in den entsprechenden da/dN-Risswachstumskurven des ASME Code, Section XI zur Bewertung des betrieblichen Verhaltens von Oberflaechenfehlern. Historisch betrachtet erfolgte in den Regelwerken die Festlegung der Vorgehensweise zur Komponentenauslegung allerdings lange vor dem gezielten experimentellen Nachweis der Umgebungseffekte auf Rissinitiierung und Risswachstum durch Ermuedung von Staehlen in Hochtemperaturwasser. Trotz dieser Erkenntnis ist es weltweit nicht zu generischen, systematischen Schaeden in Medium fuehrenden Systemen von Leichtwasserreaktoren (LWR) durch Korrosionsermuedung infolge von Fehlauslegung gekommen. Vereinzelt aufgetretene Schaeden mit deutlichen Merkmalen umgebungsbeeinflusster Ermuedungsvorgaenge konnten immer eindeutig auf das Vorkommen von unerwarteten, nicht im spezifizierten Belastungskollektiv enthaltenen Betriebstransienten zurueckgefuehrt werden. Zu diesen Ursachen zaehlen z.B. das Auftreten von thermischer Schichtung oder lokale, stroemungsinduzierte Vibrationen. In diesem Beitrag werden Ueberlegungen vorgestellt, welche die zu beobachtende Diskrepanz zwischen der diesbezueglich weltweit ueberwiegend positiven Betriebserfahrung und den Ergebnissen aus Laborversuchen mit dem

  4. 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.

  5. 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).

  6. Numerical simulation of boron injection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

    2010-02-15

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

  7. An Advanced Option for Sodium Cooled TRU Burner Loaded with Uranium-Free Fuels

    Energy Technology Data Exchange (ETDEWEB)

    You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The sodium cooled fast reactors of this kind that are called burners are designed to have low conversion ratio by reducing fuel volume fraction or reducing neutron leakage or increasing neutron absorption. However, the typical SFR burners have a limited ability of TRU burning rate due to the fact that they use metallic or oxide fuels containing fertile nuclides such as {sup 238}U and {sup 232}Th and these fertile nuclides generate fissile nuclides through neutron capture even if they are designed to have low conversion ratio (e.g., 0.6). To further enhance the TRU burning rate, the removal of the fertile nuclides from the initial fuels is required and it will accelerate the reduction of TRUs that are accumulated in storages of LWR spent fuels. However, it has been well-known 4 that the removals of the fertile nuclides from the fuel degrade the inherent safety of the SFR burner cores through the significant decrease of the fuel Doppler effect, the increase of sodium void reactivity worth, and reduction of delayed neutron fraction. In this work, new option for the sodium cooled fast TRU burner cores loaded with fertile-free metallic fuels was proposed and the new cores were designed by using the suggested option. The cores were designed to enhance the inherent safety characteristics by using axially central absorber region and 6 or 12 ZrH1.8 moderator rods per fuel assembly. For each option, we considered two different types of fertile-free ternary metallic fuel (i.e., TRU-W-10Zr and TRU-Ni-10Zr). Also, we performed the BOR (Balance of Reactivity) analyses to show the self-controllability under ATWS as a measure of inherent safety. The core performance analysis showed that the new cores using axially central absorber region substantially improve the core performance parameters such as burnup reactivity swing and sodium void reactivity worth.

  8. 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)

  9. Development of Safety Analysis Technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. B.; Kwon, Y. M.; Kim, E. K. [KAERI, Daejeon (Korea, Republic of)

    2007-06-15

    In the safety analysis code system development area, the development of an analysis code for a flow blockage could be brought to completion throughout an integrated validation of MATRA-LMR-FB. The safety analysis code of SSC-K has been evolved by building detailed reactivity models and a core 3 dimensional T/H model into it, and developing its window version. A basic analysis module for SFR features also have been developed incorporating a numerical method, best estimated correlations, and a code structure module. For the analysis of the HCDA initiating phase, a sodium boiling model to be linked to SSC-K and a fuel transient performance/cladding failure model have been developed with a state-of-the-art study on the molten fuel movement models. Besides, scoping analysis models for the post-accident heat removal phase have been developed as well. In safety analysis area, the safety criteria for the KALIMER-600 have been set up, and an internal flow channel blockage and local faults have been analyzed for the assembly safety evaluation, while key safety concepts of the KALIMER-600 has been investigated getting through the analyses of ATWS as well as design basis accidents like TOP and LOF, from which the inherent safety due to a core reactivity feedback has been assessed. The HCDA analysis for the initiating phase and an estimation of the core energy release, subsequently, have been followed with setup of the safety criteria as well as T/H analysis for the core catcher. The thermal-hydraulic behaviors, and released radioactivity sources and dose rates in the containment have been analyzed for its performance evaluation in this area. The display of a data base for research products on the KALIMER Website and the detailed process planning with its status analysis, have become feasible from achievements in the area of the integrated technology development and establishment

  10. Inherent safety analysis of the KALIMER under a LOFA with a reduced primary pump halving time

    Energy Technology Data Exchange (ETDEWEB)

    Chang, W. P.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-02-15

    The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K

  11. 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)

  12. 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.

  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. DCS Cabinet Power Loss Analysis for CPR 1000 Nuclear Power Plant%CPR1000核电厂DCS机柜失电分析

    Institute of Scientific and Technical Information of China (English)

    周亮; 赵岩峰; 孙永滨

    2014-01-01

    The DCS overall structure of CRP1000 nuclear power plant was introduced . Based on the RPC ,the signal interface character and signal processing mechanism on the key root were analyzed .By the power loss analyzing of RPC ,the RPC loss power may lead reactor trip signal from anticipated transient without scram (ATWS) system .The results indicate that it is necessary to search DCS cabinet power loss analysis .Optimi‐zing and assigning the main waterflow signals can avoid trigger reactor trip signal by mistake .The DCS cabinet power loss analysis can optimize the I&C (instrumentation and control) design and increase the nuclear plant’s reliability .%介绍了CPR1000核电厂数字化控制系统(DCS)的总体结构,以反应堆保护机柜(RPC)为基础,分析RPC的信号接口特性和信号关键路径节点的信号处理机制。结合RPC Ⅳ保护通道失电造成未能停堆的预期瞬变(ATWS)系统误发停堆信号的原因进行分析及优化,结果表明:对DCS机柜失电分析的研究是必要的,通过对RPC Ⅳ的给水流量信息进行优化和合理分配,可避免误发停堆信号。失电分析可优化仪控的设计,提高核电厂的可靠性。

  15. 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

  16. 氧化鱼油诱导草鱼 Ctenopharyngodon idellus 肝胰脏谷胱甘肽/谷胱甘肽转移酶通路的应激反应%Induced Stress of Glutathione/Glutathione Transferase Gene Pathway in Hepatopancreas of Grass Carp Ctenopharyngodon idellus by Dietary Oxidized Fish Oil

    Institute of Scientific and Technical Information of China (English)

    林秀秀; 叶元土; 蔡春芳; 吴萍; 黄雨薇; 陈科全; 李婷; 罗其刚

    2015-01-01

    Juvenile grasscarp Ctenopharyngodon idellus w ith body w eightof (74.8±1)g w ere reared in netcagesdisposed in a pond and fed five isonitrogenous and isoenergetic sem i-purified diets containing soybean oil (6S),fish oil (6F),2% oxidized fish oil (2O F),4% oxidized fish oil(4O F),and 6% oxidized fish oil(6O F)atw atertem perature of25~33℃ for72 days to study the effects ofdietary oxidized fish oilon oxidative stress pathw ay by detecting ofG SH /G STs pathw ay gene expression,G SH content,and SO D activity in hepatopancr eas of grass carp.The results show ed thatthe expression of GCLC in 6O F group w as significantly reduced (P0.05).The expression of GSR w as found to be reduced in each group,w ith significantdifference betw een 6F group and 4O F group (P0.05);各试验组 GSR 的表达活性均下调,6F 和4OF 组间显著下调(P<0.05);GSTPI 的表达活性均显著下调(P<0.05),与饲料中(EPA+DHA)含量呈线性负相关关系;除6F 组外,其余各组 MGST1的表达活性均较6S 组显著下调(P<0.05),且 MGST1的表达活性与饲料丙二醛(MDA)含量呈二项式关系,与6S 组相比,其余各组肝胰脏中 GSH 含量及 SOD 活性均显著下调(P<0.05)。氧化鱼油引起草鱼 GSH/GST 合成通路基因表达相适应,肝胰脏 GSH 合成相关基因和 MGST1的表达活性下调,而 GSTPI 的表达活性增强。GSH/GSTs 通路基因表达活性和 GSH 含量随饲料中氧化鱼油的增加呈梯度变化。

  17. 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

  18. 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

  19. 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

  20. 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

  1. Thermal Power Of The TS-300B Refrigerator in the Aspects of Statistical Research / Moc Cieplna Chłodziarki TS-300B W Aspekcie Badań Statystycznych

    Science.gov (United States)

    Nowak, Bernard; Łuczak, Rafał

    2015-09-01

    refrigerator, both on the test stand in the manufacturer's laboratory and in the workings of underground mines. The evaluation of the measurement data distributions, as well as an analysis of the basic descriptive statistics of the mentioned variables were carried out, determining their measures of central tendency, location, dispersion and asymmetry. Artykuł dotyczy poprawy cieplnych warunków pracy w wyrobiskach górniczych kopalń podziemnych stosujących lokalne systemy chłodnicze. Rozważa się w nim skuteczność schładzania powietrza chłodziarką sprężarkową bezpośredniego działania typu TS-300B. Bardzo często, w wyniku niedotrzymania wymaganych warunków pracy wymienionego systemu chłodzenia powietrza, występują rozbieżności między prognozowanymi, a więc oczekiwanymi efektami jego pracy a rzeczywistością. Dlatego, dla poprawy skuteczności pracy tego systemu, opracowano, pod kątem efektywnego wykorzystania mocy chłodniczej parownika takiej chłodziarki, łatwe w zastosowaniu praktycznym kryteria jakości. Otrzymano je w postaci modeli statystycznych określających wpływ zmiennych niezależnych, tj. parametrów powietrza wlotowego do parownika (temperatury, wilgotności i wydatku objętościowego) oraz parametrów wody chłodzącej skraplacz (temperatury i wydatku objętościowego) na moc cieplną chłodnicy powietrza traktowaną jako zmienna zależna. Równania statystyczne opisujące pracę rozważanego systemu chłodzenia powietrza wyznaczono na podstawie wielorakiej regresji liniowej i nieliniowej. Utworzone funkcje zmodyfikowano poprzez zmianę wartości współczynników w przypadku regresji liniowej oraz współczynników i wykładników w przypadku regresji nieliniowej, przy zmiennych niezależnych. Otrzymano w ten sposób funkcje dogodniejsze w praktycznych wykorzystaniach. Korzystając z metod statystyki klasycznej oceniono jakość dopasowania funkcji regresji do danych eksperymentalnych. Porównano także wartości mocy cieplnych