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Sample records for atws

  1. ATW neutronics design studies.

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-10

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

  2. Trace Assessment for BWR ATWS Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-22

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

  3. 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.

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, G.P.

    1991-01-01

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

  7. 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.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-15

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

  10. 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.

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

    DEFF Research Database (Denmark)

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

    2012-01-01

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

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

    DEFF Research Database (Denmark)

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

    2012-01-01

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

  13. 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).

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

    DEFF Research Database (Denmark)

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

    2013-01-01

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

  15. Nuclear data review and compilation for ATW systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-01-01

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

  17. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

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

    Science.gov (United States)

    2013-01-28

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

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

    Science.gov (United States)

    1998-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

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

    DEFF Research Database (Denmark)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-07

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

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

    DEFF Research Database (Denmark)

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

    2012-01-01

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

  5. Thin-shell wormholes in neo-Newtonian theory

    Science.gov (United States)

    Övgün, Ali; Salako, Ines G.

    2017-07-01

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

  6. American Veterinary Medical Association

    Science.gov (United States)

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

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

    Data.gov (United States)

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

  8. 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.

  9. Browse Title Index

    African Journals Online (AJOL)

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

  10. 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.

  11. 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.

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

    Science.gov (United States)

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

    2015-06-01

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

  13. 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.

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

    NARCIS (Netherlands)

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

    2011-01-01

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

  15. 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

  16. 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

  17. In silico

    Science.gov (United States)

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

    2018-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Laidler, J. J.

    2000-02-17

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

  19. 2011 annual meeting on nuclear technology. Section reports. Pt. 5; Jahrestagung Kerntechnik 2011. Sektionsberichte. T. 5

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-15

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

  1. 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.

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

    Science.gov (United States)

    1985-09-01

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

  3. 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...

  4. 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.)

  5. Low Repetition Rate Copper Vapor Laser.

    Science.gov (United States)

    1977-09-01

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

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

    Science.gov (United States)

    2016-10-01

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

  7. 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.

  8. Accelerator-driven Transmutation of Waste

    Science.gov (United States)

    Venneri, Francesco

    1998-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

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

    Science.gov (United States)

    1981-04-30

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

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

    Science.gov (United States)

    2006-01-01

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

  13. On Reversible Transformations of Space Elements,

    Science.gov (United States)

    1979-09-01

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

  14. BWR Anticipated Transients Without Scram Leading to Instability

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-10

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  16. 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.

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

    Directory of Open Access Journals (Sweden)

    H. Yamano

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  19. 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

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

    Science.gov (United States)

    Gudowski, Waclaw

    2000-01-01

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

  1. 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

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  5. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

  7. 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.

  8. 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.

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-08

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

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

    Science.gov (United States)

    Pak, Chan-gi; Lung, Shun-fat

    2009-01-01

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

  17. 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.

  18. Czynniki prognostyczne progresji radiologicznej w reumatoidalnym zapaleniu stawów

    Directory of Open Access Journals (Sweden)

    Piotr Wiland

    2010-08-01

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

  19. 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.

  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. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. High order boron transport scheme in TRAC-BF1

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-15

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

  8. 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.

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

    Science.gov (United States)

    Butler, Brooke

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  12. 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.

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

    Science.gov (United States)

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

    2002-11-01

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

  14. 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.

  15. Masaż Shantala – charakterystyka i spos ób wykonania

    Directory of Open Access Journals (Sweden)

    Iwona Wilk

    2015-10-01

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

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  18. 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.

  19. 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)

  20. 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)

  1. 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.

  2. 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).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  4. 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

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

    Science.gov (United States)

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

    2012-08-01

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

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ranta-Puska, Kari

    1999-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-08-15

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

  10. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  12. 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

  13. 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