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Sample records for preliminary design calculations

  1. Preliminary core design calculations for the ACPR Upgrade

    International Nuclear Information System (INIS)

    Pickard, P.S.

    1976-01-01

    The goal of the Annular Core Pulse Reactor (ACPR) Upgrade design studies is to define a core configuration that provides a significant increase in pulse fluence and fission energy deposition. The reactor modification should provide as flat an energy deposition profile for experiments as feasible. The fuels examined in this study were UO 2 -BeO (5-15 w/o UO 2 ), UC-ZrC-C (200-500 mg U/cc) and U-ZrH 1.5 . The basic core concept examined was a two region core, - a high heat capacity inner core region surrounded by an outer U-ZrH 1.5 region. Survey core calculations utilizing 1D transport calculations and cross sections libraries derived from the ORNL-AMPX code examined relative fuel loadings, fuel temperatures, reactivity requirements and pulse performance improvement. Reference designs for all candidate fuels were defined utilizing 2D transport and Monte Carlo calculations. The performance implications of alternative core designs were also examined for the UO 2 -BeO and UC-ZrC-C fuel candidates. (author)

  2. Preliminary Calculation for Plasma Chamber Design of Pulsed Electron Source Based on Plasma

    International Nuclear Information System (INIS)

    Widdi Usada

    2009-01-01

    This paper described the characteristics of pulsed electron sources with anode-cathode distance of 5 cm, electrode diameter of 10 cm, driven by capacitor energy of 25 J. The preliminary results showed that if the system is operated with diode resistance is 1.6 Ω, plasma resistance is 0.14 Ω, and β is 0.94, the achieved of plasma voltage is 640 V, its current is 4.395 kA with its pulse width of 0.8 μsecond. According to breakdown voltage based on Paschen empirical formula, with this achieved voltage, this system could be operated for operation pressure of 1 torr. (author)

  3. Proposal of a calculation methodology for the preliminary design of a coalescing filter

    International Nuclear Information System (INIS)

    Gonzalez Dobrosky, Cintia

    2015-01-01

    Coalescing filters are described which are equipments for capture and recovery of mist most efficient, inexpensive and have fewer limitations of application. The operation, equations and ideal characteristics of filter media of these models are explained. A methodology for design and scale-up of this type of equipment for liquid recovery in gaseous currents is proposed from experimental tests, in order to guide the interested reader in its making. (author) [es

  4. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  5. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

    International Nuclear Information System (INIS)

    Siefken, L.J.

    1999-01-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

  6. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  7. OMEGA Upgrade preliminary design

    International Nuclear Information System (INIS)

    Craxton, R.S.

    1989-10-01

    The OMEGA laser system at the Laboratory for Laser Energetics of the University of Rochester is the only major facility in the United States capable of conducting fully diagnosed, direct-drive, spherical implosion experiments. As such, it serves as the national Laser Users Facility, benefiting scientists throughout the country. The University's participation in the National Inertial Confinement Fusion (ICF) program underwent review by a group of experts under the auspices of the National Academy of Sciences (the Happer Committee) in 1985. The Happer Committee recommended that the OMEGA laser be upgraded in energy to 30 kJ. To this end, Congress appropriated $4,000,000 for the preliminary design of the OMEGA Upgrade, spread across FY88 and FY89. This document describes the preliminary design of the OMEGA Upgrade. The proposed enhancements to the existing OMEGA facility will result in a 30-kHJ, 351-nm, 60-beam direct-drive system, with a versatile pulse-shaping facility and a 1%--2% uniformity of target drive. The Upgrade will allow scientists to explore the ignition-scaling regime, and to study target behavior that is hydrodynamically equivalent to that of targets appropriate for a laboratory microfusion facility (LMF). In addition, it will be possible to perform critical interaction experiments with large-scale-length uniformly irradiated plasmas

  8. HEU benchmark calculations and LEU preliminary calculations for IRR-1

    International Nuclear Information System (INIS)

    Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.

    2004-01-01

    We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

  9. Preliminary designs: passive solar manufactured housing. Technical status report

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-12

    The criteria established to guide the development of the preliminary designs are listed. Three preliminary designs incorporating direct gain and/or sunspace are presented. Costs, drawings, and supporting calculations are included. (MHR)

  10. Preliminary isodose calculation for gynecological curietherapy

    International Nuclear Information System (INIS)

    Bridier, A.; Dutreix, A.; Gerbaulet, A.; Chassagne, D.

    1981-01-01

    We present a preliminary method of calculating the dimensions of the reference isodose, based upon the geometrical distribution and length of the sources used, their linear activity and the length of treatment, that does not require use of a computer. Inversely, this method can be used to determine the factors necessary to produce a given shape of isodose, and also to predict the change in shape of the isodose that will be produced by altering the various factors. This method was derived from a systematic computer study of dose distribution in which each factor was varied independently of all others. The dimensions of the isodoses, calculated by this method, were found to be in agreement with those derived from computer calculation to within an error of about 2 mm. The method is only applicable for a limited range of positions of the vaginal sources. The influence of the positions of these sources along the line of the axis of uterine catheter and of their inclination to this line, are currently being studied. The results are presented as mathematical formulae relating each dimension of the isodose curves to the features of the application, but could equally well be expressed in tabular form that would be more convenient for everyday use. An example of the calculation used is given to facilitate understanding of the method [fr

  11. Preliminary design county plan Zeeland

    International Nuclear Information System (INIS)

    1987-01-01

    The preliminary design 'Streekplan Zeeland' (Country plan Zeeland, with regard to the location of additional nuclear power plants in Zeeland, the Netherlands) has passed through a consultation and participation round. Thereupon 132 reactions have been received. These have been incorporated and answered in two notes. This proposal deals with the principal points of the preliminary design and treats also the remarks of the committees Environmental (town and country) Planning (RO), Provincial (town and country) Planning Committee (PPC) and Association of Communities of Zeeland (VZG), on the reply notes. The preliminary design with the modifications, collected in appendix 3, is proposed to be the starting point in the drawing-up of the design-country-plan. This design subsequently will pass the formal country-plan procedure. (author). 1 fig

  12. Space reactor preliminary mechanical design

    International Nuclear Information System (INIS)

    Meier, K.L.

    1983-01-01

    An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO 2 block core mounts, bolted collar fuel module restraints, and a BeO central plug

  13. Preliminary PBFA II design

    International Nuclear Information System (INIS)

    Johnson, D.L.; VanDevender, J.P.; Martin, T.H.

    1980-01-01

    The upgrade of Sandia National Laboratories particle beam fusion accelerator, PBFA I, to PBFA II presents several interesting and challenging pulsed power design problems. PBFA II requires increasing the PBFA I output parameters from 2 MV, 30 TW, 1 MJ to 4 MV, 100 TW, 3.5 MJ with the constraint of using much of the same PBFA I hardware. The increased PBFA II output will be obtained by doubling the number of modules (from 36 to 72), increasing the primary energy storage (from 4 MJ to 15 MJ), lowering the pulse forming line (PFL) output impedance, and adding a voltage doubling network

  14. KALIMER preliminary conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

    2000-08-01

    This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report.

  15. KALIMER preliminary conceptual design report

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

    2000-08-01

    This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report

  16. Versator divertor experiment: preliminary designs

    International Nuclear Information System (INIS)

    Wan, A.S.; Yang, T.F.

    1984-08-01

    The emergence of magnetic divertors as an impurity control and ash removal mechanism for future tokamak reactors bring on the need for further experimental verification of the divertor merits and their ability to operate at reactor relevant conditions, such as with auxiliary heating. This paper presents preliminary designs of a bundle and a poloidal divertor for Versator II, which can operate in conjunction with the existing 150 kW of LHRF heating or LH current drive. The bundle divertor option also features a new divertor configuration which should improve the engineering and physics results of the DITE experiment. Further design optimization in both physics and engineering designs are currently under way

  17. BIPS-FS preliminary design, miscellaneous notes

    International Nuclear Information System (INIS)

    1976-01-01

    A compendium of flight system preliminary design internal memos and progress report extracts for the Brayton Isotope Power System Preliminary Design Review to be held July 20, 21, and 22, 1975 is presented. The purpose is to bring together those published items which relate only to the preliminary design of the Flight System, Task 2 of Phase I. This preliminary design effort was required to ensure that the Ground Demonstration System will represent the Flight System as closely as possible

  18. Exploratory shaft facility preliminary designs - Permian Basin

    International Nuclear Information System (INIS)

    1983-09-01

    The purpose of the Preliminary Design Report, Permian Basin, is to provide a description of the preliminary design for an Exploratory Shaft Facility in the Permian Basin, Texas. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers are included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description, and Construction Cost Estimate. 30 references, 13 tables

  19. Closure and Sealing Design Calculation

    International Nuclear Information System (INIS)

    T. Lahnalampi; J. Case

    2005-01-01

    The purpose of the ''Closure and Sealing Design Calculation'' is to illustrate closure and sealing methods for sealing shafts, ramps, and identify boreholes that require sealing in order to limit the potential of water infiltration. In addition, this calculation will provide a description of the magma that can reduce the consequences of an igneous event intersecting the repository. This calculation will also include a listing of the project requirements related to closure and sealing. The scope of this calculation is to: summarize applicable project requirements and codes relating to backfilling nonemplacement openings, removal of uncommitted materials from the subsurface, installation of drip shields, and erecting monuments; compile an inventory of boreholes that are found in the area of the subsurface repository; describe the magma bulkhead feature and location; and include figures for the proposed shaft and ramp seals. The objective of this calculation is to: categorize the boreholes for sealing by depth and proximity to the subsurface repository; develop drawing figures which show the location and geometry for the magma bulkhead; include the shaft seal figures and a proposed construction sequence; and include the ramp seal figure and a proposed construction sequence. The intent of this closure and sealing calculation is to support the License Application by providing a description of the closure and sealing methods for the Safety Analysis Report. The closure and sealing calculation will also provide input for Post Closure Activities by describing the location of the magma bulkhead. This calculation is limited to describing the final configuration of the sealing and backfill systems for the underground area. The methods and procedures used to place the backfill and remove uncommitted materials (such as concrete) from the repository and detailed design of the magma bulkhead will be the subject of separate analyses or calculations. Post-closure monitoring will not

  20. Ship design methodologies of preliminary design

    CERN Document Server

    Papanikolaou, Apostolos

    2014-01-01

    This book deals with ship design and in particular with methodologies of the preliminary design of ships. The book is complemented by a basic bibliography and five appendices with useful updated charts for the selection of the main dimensions and other basic characteristics of different types of ships (Appendix A), the determination of hull form  from the data of systematic hull form series (Appendix B), the detailed description of the relational method for the preliminary estimation of ship weights (Appendix C), a brief review of the historical evolution of shipbuilding science and technology from the prehistoric era to date (Appendix D) and finally a historical review of regulatory developments of ship's damage stability to date (Appendix E).  The book can be used as textbook for ship design courses or as additional reading for university or college students of naval architecture courses and related disciplines; it may also serve as a reference book for naval architects, practicing engineers of rel...

  1. Preliminary design of smart fuel

    International Nuclear Information System (INIS)

    Kim, Y.; Ha, D.; Park, S.; Nahm, K.; Lee, K.; Kim, J.

    2007-01-01

    SMART (System-integrated Modular Advanced Reactor) is a novel light water rector with a modular, integral primary system configuration. This concept has been developing a 660 MWt by Korean Nuclear Power Industry Group with KAERI. SMART is being developed for use as an energy source for small-scale power generation and seawater desalination. Although the design of SMART is based on the current pressurized water reactor technology, new technologies such as enhanced safety, and passive safety have been applied, and system simplification and modularization, innovations in manufacturing and installation technologies have been implemented culminating in a design that has enhanced safety and economy, and is environment -friendly. In this paper described the preliminary design of the nuclear Fuel for this SMART, the design concept and the characteristics of SMART Fuel. In specially this paper describe the optimization of grid span adjustment to improve the thermal performance of the SMART Fuel as well as to improve the seismic resistance performance of the SMART Fuel, it is not easy to improve the both performance simultaneously because of design parameter of each performance inversely proportional. SMART Fuel enable to extra-long extended fuel cycle length and resistance of proliferation, enhanced safety, improved economics and reduced nuclear waste

  2. Electronics reliability calculation and design

    CERN Document Server

    Dummer, Geoffrey W A; Hiller, N

    1966-01-01

    Electronics Reliability-Calculation and Design provides an introduction to the fundamental concepts of reliability. The increasing complexity of electronic equipment has made problems in designing and manufacturing a reliable product more and more difficult. Specific techniques have been developed that enable designers to integrate reliability into their products, and reliability has become a science in its own right. The book begins with a discussion of basic mathematical and statistical concepts, including arithmetic mean, frequency distribution, median and mode, scatter or dispersion of mea

  3. Preliminary design data package. Appendix C

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-25

    The design requirements, design philosophy, method and assumptions, and preliminary computer-aided design of the Near-Term Hybrid Vehicle including its electric and heat power units, control equipment, transmission system, body, and overall vehicle characteristics are presented. (LCL)

  4. Calculable resistors of coaxial design

    International Nuclear Information System (INIS)

    Kucera, J; Vollmer, E; Schurr, J; Bohacek, J

    2009-01-01

    1000 Ω and 1290.64 Ω coaxial resistors with calculable frequency dependence have been realized at PTB to be used in quantum Hall effect-based impedance measurements. In contradistinction to common designs of coaxial resistors, the design described in this paper makes it possible to remove the resistive element from the shield and to handle it without cutting the outer cylindrical shield of the resistor. Emphasis has been given to manufacturing technology and suppressing unwanted sources of frequency dependence. The adjustment accuracy is better than 10 µΩ Ω −1

  5. Preliminary design of RDE feedwater pump impeller

    International Nuclear Information System (INIS)

    Sri Sudadiyo

    2018-01-01

    Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Experimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. (author)

  6. Nuclear data library in design calculation

    International Nuclear Information System (INIS)

    Hirano, Go; Kosaka, Shinya

    2006-01-01

    In core design calculation, nuclear data takes part as multi group cross section library during the assembly calculation, which is the first stage of a core design calculation. This report summarizes the multi group cross section libraries used in assembly calculations and also presents the methods adopted for resonance and assembly calculation. (author)

  7. Preliminary Analysis For Wolsong Par Effects Using ISACC Calculations

    International Nuclear Information System (INIS)

    Song, Yong Mann; Kim, Dong Ha

    2012-01-01

    In the paper, hydrogen control effects using PARs only are analyzed for severe SBO station blackout (SBO) sequences beyond the design basis accidents in WS-1 which are of CANDU6 type reactor. As a computational tool, the latest version of ISAAC4.3 (Integrated Severe Accident Analysis Code for CANDU), which is a fully integrated and lumped severe accident computer code, is used to simulate hydrogen generation and transport inside the reactor building (R/B) before its failure. For the performance of hydrogen removal, the depletion rate equation of K-PAR developed in Korea is applied. In a CANDU reactor, three areas are identified as sources of hydrogen under severe accidents: fuel-coolant interactions in intact channels, suspended fuel or debris interactions in-calandria tank and debris interactions in-calandria vault. The first two origins provide source for the late ('late' terminology is used because it takes more than one day before calandria tank failure) potential hydrogen combustion before calandria tank failure and all the three origins would provide source for the very late potential hydrogen combustion occurring at or after calaria tank failure. If the hydrogen mitigation system fails, the AICC (adiabatic isochoric complete combustion) burning of highly flammable hydrogen may cause Wolsong R/B failure. So hydrogen induced failure possibility is evaluated, using preliminary ISAAC calculations, under several SBO conditions with and without PAR for both late and very late accident periods

  8. Preliminary topical report on comparison reactor disassembly calculations

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1975-11-01

    Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2-POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherent in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident

  9. Preliminary physics calculations for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Kalimullah.

    1975-01-01

    Calculations of sodium void, fuel, and clad worths, power distribution, and control rod worths have been carried out for an R-Z model of the CRBR, using diffusion theory and first-order perturbation theory for material worths. The power distribution and control rod worths have also been calculated in two-dimensional triangular mesh geometry. The present results are preliminary because of inaccuracy of the reactor model and the cross sections used, but the final results are not expected to be greatly different. (U.S.)

  10. Preliminary I&C Design for LORELEI

    International Nuclear Information System (INIS)

    Korotkin, S.; Kaufman, Y.; Guttmann, E. B.; Levy, S.; Amidan, D.; Gdalyho, B.; Cahana, T.; Ellenbogen, A.; Arad, M.; Weiss, Y.; Sasson, A.; Ferry, L.; Bourrelly, F.; Cohen, Y.

    2014-01-01

    This document summarizes the preliminary I&C design for LORELEI experiment The preliminary design deals with considerations regarding appropriate safety and service instrumentation. The determined closed loop control rules for temperature and position will be implemented in the detailed design. The Computer Aided Operator Decisions System (CAODS) will be used for prediction of hot spot temperature and thickness of oxidation layer using Baker-Just correlation. The proposed hybrid simulation system comprising of both virtual and real hardware will be in-cooperated for LORELEI verification. It will perform both integration cold tests for a partial hardware loop and virtual tests for the final I&C design

  11. Nuclear Characteristics of SPNDs and Preliminary Calculation of Hybrid Fixed Incore Detector with Monte Carlo Code

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Lee, Kyung Hoon; Song, Jae Seung; Park, Sang Yoon

    2013-01-01

    In this paper, the basic nuclear characteristics of major emitter materials were surveyed. In addition, preliminary calculations of Cobalt-Vanadium fixed incore detector were performed using the Monte Carlo code. Calculational results were cross-checked by KARMA. KARMA is a two-dimensional multigroup transport theory code developed by the KAERI and approved by Korean regularity agency to be employed as a nuclear design tool for a Korean commercial pressurizer water reactor. The nuclear characteristics of the major emitter materials were surveyed, and preliminary calculations of the hybrid fixed incore detector were performed with the MCNP code. The eigenvalue and pin-by-pin fission power distributions were calculated and showed good agreement with the KARMA calculation results. As future work, gamma power distributions as well as several types of XS of the emitter, insulator, and collector regions for a Co-V ICI assembly will be evaluated and compared

  12. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

  13. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  14. Preliminary Design of a LSA Aircraft Using Wind Tunnel Tests

    Directory of Open Access Journals (Sweden)

    Norbert ANGI

    2015-12-01

    Full Text Available This paper presents preliminary results concerning the design and aerodynamic calculations of a light sport aircraft (LSA. These were performed for a new lightweight, low cost, low fuel consumption and long-range aircraft. The design process was based on specific software tools as Advanced Aircraft Analysis (AAA, XFlr 5 aerodynamic and dynamic stability analysis, and Catia design, according to CS-LSA requirements. The calculations were accomplished by a series of tests performed in the wind tunnel in order to assess experimentally the aerodynamic characteristics of the airplane.

  15. Life cycle analysis in preliminary design stages

    OpenAIRE

    Agudelo , Lina-Maria; Mejía-Gutiérrez , Ricardo; Nadeau , Jean-Pierre; PAILHES , Jérôme

    2014-01-01

    International audience; In a design process the product is decomposed into systems along the disciplinary lines. Each stage has its own goals and constraints that must be satisfied and has control over a subset of design variables that describe the overall system. When using different tools to initiate a product life cycle, including the environment and impacts, its noticeable that there is a gap in tools that linked the stages of preliminary design and the stages of materialization. Differen...

  16. Preliminary conceptual design and analysis on KALIMER reactor structures

    International Nuclear Information System (INIS)

    Kim, Jong Bum

    1996-10-01

    The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

  17. HTGR gas turbine power plant preliminary design

    International Nuclear Information System (INIS)

    Koutz, S.L.; Krase, J.M.; Meyer, L.

    1973-01-01

    The preliminary reference design of the HTGR gas turbine power plant is presented. Economic and practical problems and incentives related to the development and introduction of this type of power plant are evaluated. The plant features and major components are described, and a discussion of its performance, economics, development, safety, control, and maintenance is presented. 4 references

  18. KALIMER fuel system preliminary design description

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B.O.; Nam, C.; Paek, S.K.

    1998-10-01

    This document provides general design concepts, design basis, preliminary design specification and design technologies which are needed for designing the fuel/non-fuel rods and assembly ducts of the KALIMER fuel system. The core of LMFBR consists of driver fuel assembly, blanket assembly, reflector assembly, shielding assembly, control assembly and GEM (Gas Expansion Module) as well as USS, dummy assembly, detector assembly. These core components must be designed to withstand the high temperature, high flux for a long irradiation exposure time. Due to the high temperature and high flux, irradiation creep and swelling as well as thermal-mechanical deformation are occurred at the fuel/non-fuel system and cause the deformations of materials and the geometric deflections at fuel/non-fuel rods, assembly ducts and components. In order to overcome these intricate phenomena through the engineering design, the design basis including theoretical analysis methodologies and design considerations, material characteristics of fuel system, and the specifications and drawings of fuel/non-fuel rods and assembly ducts, respectively, are presented. This document is preliminary design description which is produced in the conceptual design stage, and does not present the detailed and finalized design data which can be for the manufacturing. (author). 22 refs

  19. Safety performance of preliminary KALIMER conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

    1999-07-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  20. Safety performance of preliminary KALIMER conceptual design

    International Nuclear Information System (INIS)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

    1999-01-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  1. Preliminary Calculations of Shutdown Dose Rate for the CTS Diagnostics System

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Nonbøl, Erik; Lauritzen, Bent

    2015-01-01

    DTU and IST 2 are partners in the design of a collective Thomson Scattering (CTS) diagnostics for ITER through a contract with F4E. The CTS diagnostic utilizes probing radiation of ~60 GHz emitted into the plasma and, using a mirror, collects the scattered radiation by an array of receivers. Having...... on supplying input which affect the system design. Examples include: - Heatloads on plasma facing mirrors and preliminary stress and thermal analysis - Port plug cooling requirements and it's dependence on system design (in particular blanket cut-out) - Shutdown dose-rate calculations (relative analysis...

  2. Preliminary core design of IRIS-50

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Franceschini, Fausto

    2009-01-01

    IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

  3. Preliminary design studies for the DESCARTES and CIDER codes

    International Nuclear Information System (INIS)

    Eslinger, P.W.; Miley, T.B.; Ouderkirk, S.J.; Nichols, W.E.

    1992-12-01

    The Hanford Environmental Dose Reconstruction (HEDR) project is developing several computer codes to model the release and transport of radionuclides into the environment. This preliminary design addresses two of these codes: Dynamic Estimates of Concentrations and Radionuclides in Terrestrial Environments (DESCARTES) and Calculation of Individual Doses from Environmental Radionuclides (CIDER). The DESCARTES code will be used to estimate the concentration of radionuclides in environmental pathways, given the output of the air transport code HATCHET. The CIDER code will use information provided by DESCARTES to estimate the dose received by an individual. This document reports on preliminary design work performed by the code development team to determine if the requirements could be met for Descartes and CIDER. The document contains three major sections: (i) a data flow diagram and discussion for DESCARTES, (ii) a data flow diagram and discussion for CIDER, and (iii) a series of brief statements regarding the design approach required to address each code requirement

  4. Preliminary design of a coffee harvester

    Directory of Open Access Journals (Sweden)

    Raphael Magalhães Gomes Moreira

    2016-10-01

    Full Text Available Design of an agricultural machine is a highly complex process due to interactions between the operator, machine, and environment. Mountain coffee plantations constitute an economic sector that requires huge investments for the development of agricultural machinery to improve the harvesting and post-harvesting processes and to overcome the scarcity of work forces in the fields. The aim of this study was to develop a preliminary design for a virtual prototype of a coffee fruit harvester. In this study, a project methodology was applied and adapted for the development of the following steps: project planning, informational design, conceptual design, and preliminary design. The construction of a morphological matrix made it possible to obtain a list of different mechanisms with specific functions. The union between these mechanisms resulted in variants, which were weighed to attribute scores for each selected criterion. From each designated proposal, two variants with the best scores were selected and this permitted the preparation of the preliminary design of both variants. The archetype was divided in two parts, namely the hydraulically articulated arms and the harvesting system that consisted of the vibration mechanism and the detachment mechanism. The proposed innovation involves the use of parallel rods, which were fixed in a plane and rectangular metal sheet. In this step, dimensions including a maximum length of 4.7 m, a minimum length of 3.3 m, and a total height of 2.15 m were identified based on the functioning of the harvester in relation to the coupling point of the tractor.

  5. Exploratory shaft facility preliminary designs - Gulf Interior Region salt domes

    International Nuclear Information System (INIS)

    1983-09-01

    The purpose of the Preliminary Design Report, Gulf Interior Region, is to provide a description of the preliminary design for an Exploratory Shaft Facility on the Richton Dome, Mississippi. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers are included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description and Construction Cost Estimate

  6. Exploratory shaft facility preliminary designs - Paradox Basin. Technical report

    International Nuclear Information System (INIS)

    1983-09-01

    The purpose of the Preliminary Design Report, Paradox Basin, is to provide a description of the preliminary design for an Exploratory Shaft Facility in the Paradox Basin, Utah. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling Method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers is included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description, and Construction Cost Estimate. 30 references

  7. Design review report for the hydrogen interlock preliminary design

    International Nuclear Information System (INIS)

    Corbett, J.E.

    1996-01-01

    This report documents the completion of a preliminary design review for the hydrogen interlock. The hydrogen interlock, a proposed addition to the Rotary Mode Core Sampling (RMCS) system portable exhauster, is intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to validate basic design assumptions and concepts to support a path forward leading to a final design. The conclusion reached by the review committee was that the design was acceptable and efforts should continue toward a final design review

  8. Business System Planning Project, Preliminary System Design

    International Nuclear Information System (INIS)

    EVOSEVICH, S.

    2000-01-01

    CH2M HILL Hanford Group, Inc. (CHG) is currently performing many core business functions including, but not limited to, work control, planning, scheduling, cost estimating, procurement, training, and human resources. Other core business functions are managed by or dependent on Project Hanford Management Contractors including, but not limited to, payroll, benefits and pension administration, inventory control, accounts payable, and records management. In addition, CHG has business relationships with its parent company CH2M HILL, U.S. Department of Energy, Office of River Protection and other River Protection Project contractors, government agencies, and vendors. The Business Systems Planning (BSP) Project, under the sponsorship of the CH2M HILL Hanford Group, Inc. Chief Information Officer (CIO), have recommended information system solutions that will support CHG business areas. The Preliminary System Design was developed using the recommendations from the Alternatives Analysis, RPP-6499, Rev 0 and will become the design base for any follow-on implementation projects. The Preliminary System Design will present a high-level system design, providing a high-level overview of the Commercial-Off-The-Shelf (COTS) modules and identify internal and external relationships. This document will not define data structures, user interface components (screens, reports, menus, etc.), business rules or processes. These in-depth activities will be accomplished at implementation planning time

  9. Heat Exchanger Support Bracket Design Calculations

    International Nuclear Information System (INIS)

    Rucinski, Russ

    1995-01-01

    This engineering note documents the design of the heat exchanger support brackets. The heat exchanger is roughly 40 feet long, 22 inches in diameter and weighs 6750 pounds. It will be mounted on two identical support brackets that are anchored to a concrete wall. The design calculations were done for one bracket supporting the full weight of the heat exchanger, rounded up to 6800 pounds. The design follows the American Institute of Steel Construction (AISC) Manual of steel construction, Eighth edition. All calculated stresses and loads on welds were below allowables.

  10. Preliminary design for a maglev development facility

    Energy Technology Data Exchange (ETDEWEB)

    Coffey, H.T.; He, J.L.; Chang, S.L.; Bouillard, J.X.; Chen, S.S.; Cai, Y.; Hoppie, L.O.; Lottes, S.A.; Rote, D.M. (Argonne National Lab., IL (United States)); Zhang, Z.Y. (Polytechnic Univ., Brooklyn, NY (United States)); Myers, G.; Cvercko, A. (Sterling Engineering, Westchester, IL (United States)); Williams, J.R. (Alfred Benesch and Co., Chicago, IL (United States))

    1992-04-01

    A preliminary design was made of a national user facility for evaluating magnetic-levitation (maglev) technologies in sizes intermediate between laboratory experiments and full-scale systems. A technical advisory committee was established and a conference was held to obtain advice on the potential requirements of operational systems and how the facility might best be configured to test these requirements. The effort included studies of multiple concepts for levitating, guiding, and propelling maglev vehicles, as well as the controls, communications, and data-acquisition and -reduction equipment that would be required in operating the facility. Preliminary designs for versatile, dual 2-MVA power supplies capable of powering attractive or repulsive systems were developed. Facility site requirements were identified. Test vehicles would be about 7.4 m (25 ft) long, would weigh form 3 to 7 metric tons, and would operate at speeds up to 67 m/s (150 mph) on a 3.3-km (2.05-mi) elevated guideway. The facility would utilize modular vehicles and guideways, permitting the substitution of levitation, propulsion, and guideway components of different designs and materials for evaluation. The vehicle would provide a test cell in which individual suspension or propulsion components or subsystems could be tested under realistic conditions. The system would allow economical evaluation of integrated systems under varying weather conditions and in realistic geometries.

  11. Preliminary safety design analysis of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Kwon, Y. M.; Kim, K. D. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This document first introduces a set of safety design requirements and accident evaluation criteria established for the conceptual design of KALIMER and then summarizes some of the preliminary results of engineering and design analyses performed for the safety of KALIMER. 19 refs., 19 figs., 6 tabs. (Author)

  12. PRELIMINARY SELECTION OF MGR DESIGN BASIS EVENTS

    International Nuclear Information System (INIS)

    Kappes, J.A.

    1999-01-01

    The purpose of this analysis is to identify the preliminary design basis events (DBEs) for consideration in the design of the Monitored Geologic Repository (MGR). For external events and natural phenomena (e.g., earthquake), the objective is to identify those initiating events that the MGR will be designed to withstand. Design criteria will ensure that radiological release scenarios resulting from these initiating events are beyond design basis (i.e., have a scenario frequency less than once per million years). For internal (i.e., human-induced and random equipment failures) events, the objective is to identify credible event sequences that result in bounding radiological releases. These sequences will be used to establish the design basis criteria for MGR structures, systems, and components (SSCs) design basis criteria in order to prevent or mitigate radiological releases. The safety strategy presented in this analysis for preventing or mitigating DBEs is based on the preclosure safety strategy outlined in ''Strategy to Mitigate Preclosure Offsite Exposure'' (CRWMS M andO 1998f). DBE analysis is necessary to provide feedback and requirements to the design process, and also to demonstrate compliance with proposed 10 CFR 63 (Dyer 1999b) requirements. DBE analysis is also required to identify and classify the SSCs that are important to safety (ITS)

  13. Preliminary Design of a Femtosecond Oscilloscope

    CERN Document Server

    Gazazyan, Edmond D; Kalantaryan, Davit K; Laziev, Edouard; Margaryan, Amour

    2005-01-01

    The calculations on motion of electrons in a finite length electromagnetic field of linearly and circularly polarized laser beams have shown that one can use the transversal deflection of electrons on a screen at a certain distance after the interaction region for the measurement of the length and longitudinal particle distribution of femtosecond bunches. In this work the construction and preliminary parameters of various parts of a device that may be called femtosecond oscilloscope are considered. The influence of various factors, such as the energy spread and size of the electron bunches, are taken into account. For CO2 laser intensity 1016 W/cm2 and field free drift length 1m the deflection is 5.3 and 0.06 cm, while the few centimeters long interaction length between 2 mirrors requires assembling accuracy 6 mm and 1.3 micron for 20 MeV to 50 keV, respectively.

  14. NSLS-II Preliminary Design Report

    International Nuclear Information System (INIS)

    Dierker, S.

    2007-01-01

    Following the CD0 approval of the National Synchrotron Light Source II (NSLS-II) during August 2005, Brookhaven National Laboratory prepared a conceptual design for a worldclass user facility for scientific research using synchrotron radiation. DOE SC review of the preliminary baseline in December 2006 led to the subsequent CD1 approval (approval of alternative selection and cost range). This report is the documentation of the preliminary design work for the NSLS-II facility. The preliminary design of the Accelerator Systems (Part 1) was developed mostly based of the Conceptual Design Report, except for the Booster design, which was changed from in-storage-ring tunnel configuration to in external- tunnel configuration. The design of beamlines (Part 2) is based on designs developed by engineering firms in accordance with the specification provided by the Project. The conventional facility design (Part 3) is the Title 1 preliminary design by the AE firm that met the NSLS-II requirements. Last and very important, Part 4 documents the ES and H design and considerations related to this preliminary design. The NSLS-II performance goals are motivated by the recognition that major advances in many important technology problems will require scientific breakthroughs in developing new materials with advanced properties. Achieving this will require the development of new tools that will enable the characterization of the atomic and electronic structure, chemical composition, and magnetic properties of materials, at nanoscale resolution. These tools must be nondestructive, to image and characterize buried structures and interfaces, and they must operate in a wide range of temperatures and harsh environments. The NSLS-II facility will provide ultra high brightness and flux and exceptional beam stability. It will also provide advanced insertion devices, optics, detectors, and robotics, and a suite of scientific instruments designed to maximize the scientific output of the

  15. NSLS-II Preliminary Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Dierker, S.

    2007-11-01

    Following the CD0 approval of the National Synchrotron Light Source II (NSLS-II) during August 2005, Brookhaven National Laboratory prepared a conceptual design for a worldclass user facility for scientific research using synchrotron radiation. DOE SC review of the preliminary baseline in December 2006 led to the subsequent CD1 approval (approval of alternative selection and cost range). This report is the documentation of the preliminary design work for the NSLS-II facility. The preliminary design of the Accelerator Systems (Part 1) was developed mostly based of the Conceptual Design Report, except for the Booster design, which was changed from in-storage-ring tunnel configuration to in external- tunnel configuration. The design of beamlines (Part 2) is based on designs developed by engineering firms in accordance with the specification provided by the Project. The conventional facility design (Part 3) is the Title 1 preliminary design by the AE firm that met the NSLS-II requirements. Last and very important, Part 4 documents the ES&H design and considerations related to this preliminary design. The NSLS-II performance goals are motivated by the recognition that major advances in many important technology problems will require scientific breakthroughs in developing new materials with advanced properties. Achieving this will require the development of new tools that will enable the characterization of the atomic and electronic structure, chemical composition, and magnetic properties of materials, at nanoscale resolution. These tools must be nondestructive, to image and characterize buried structures and interfaces, and they must operate in a wide range of temperatures and harsh environments. The NSLS-II facility will provide ultra high brightness and flux and exceptional beam stability. It will also provide advanced insertion devices, optics, detectors, and robotics, and a suite of scientific instruments designed to maximize the scientific output of the facility

  16. Site Characterization and Preliminary Performance Assessment Calculation Applied To JAEA-Horonobe URL Site of Japan

    International Nuclear Information System (INIS)

    Lim, Doo Hyun; Hatanaka, Koichiro; Ishii, Eiichi

    2010-01-01

    JAEA-Horonobe Underground Research Laboratory (URL) is designed for research and development on high-level radioactive waste (HLW) repository in sedimentary rock. For a potential HLW repository, understanding and implementing fracturing and faulting system, with data from the site characterization, into the performance assessment is essential because fracture and fault will be the major conductors or barriers for the groundwater flow and radionuclide release. The objectives are i) quantitative derivation of characteristics and correlation of fracturing/faulting system with geologic and geophysics data obtained from the site characterization, and ii) preliminary performance assessment calculation with characterized site information

  17. Site Characterization and Preliminary Performance Assessment Calculation Applied To JAEA-Horonobe URL Site of Japan

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Doo Hyun [NE Union Hill Road, Suite 200, WA 98052 (United States); Hatanaka, Koichiro; Ishii, Eiichi [Japan Atomic Energy Agency, Hokkaido (Japan)

    2010-10-15

    JAEA-Horonobe Underground Research Laboratory (URL) is designed for research and development on high-level radioactive waste (HLW) repository in sedimentary rock. For a potential HLW repository, understanding and implementing fracturing and faulting system, with data from the site characterization, into the performance assessment is essential because fracture and fault will be the major conductors or barriers for the groundwater flow and radionuclide release. The objectives are i) quantitative derivation of characteristics and correlation of fracturing/faulting system with geologic and geophysics data obtained from the site characterization, and ii) preliminary performance assessment calculation with characterized site information

  18. IRIS: Proceeding Towards the Preliminary Design

    International Nuclear Information System (INIS)

    Carelli, M.; Miller, K.; Lombardi, C.; Todreas, N.; Greenspan, E.; Ninokata, H.; Lopez, F.; Cinotti, L.; Collado, J.; Oriolo, F.; Alonso, G.; Morales, M.; Boroughs, R.; Barroso, A.; Ingersoll, D.; Cavlina, N.

    2002-01-01

    The IRIS (International Reactor Innovative and Secure) project has completed the conceptual design phase and is moving towards completion of the preliminary design, scheduled for the end of 2002. Several other papers presented in this conference provide details on major aspects of the IRIS design. The three most innovative features which uniquely characterize IRIS are, in descending order of impact: 1. Safety-by-design, which takes maximum advantage of the integral configuration to eliminate from consideration some accidents, greatly lessen the consequence of other accident scenarios and decrease their probability of occurring; 2. Optimized maintenance, where the interval between maintenance shutdowns is extended to 48 months; and 3. Long core life, of at least four years without shuffling or partial refueling. Regarding feature 1, design and analyses will be supplemented by an extensive testing campaign to verify and demonstrate the performance of the integral components, individually as well as interactive systems. Test planning is being initiated. Test results will be factored into PRA analyses under an overall risk informed regulation approach, which is planned to be used in the IRIS licensing. Pre-application activities with NRC are also scheduled to start in mid 2002. Regarding feature 2, effort is being focused on advanced online diagnostics for the integral components, first of all the steam generators, which are the most critical component; several techniques are being investigated. Finally, a four year long life core design is well underway and some of the IRIS team members are examining higher enrichment, eight to ten year life cores which could be considered for reloads. (authors)

  19. Preliminary Opto-Mechanical Design for the X2000 Transceiver

    Science.gov (United States)

    Hemmati, H.; Page, N. A.

    2000-01-01

    Preliminary optical design and mechanical conceptual design for a 30 cm aperture transceiver are described. A common aperture is used for both transmit and receive. Special attention was given to off-axis and scattered light rejection and isolation of the receive channel from the transmit channel. Requirements, details of the design and preliminary performance analysis of the transceiver are provided.

  20. Monte Carlo methods for shield design calculations

    International Nuclear Information System (INIS)

    Grimstone, M.J.

    1974-01-01

    A suite of Monte Carlo codes is being developed for use on a routine basis in commercial reactor shield design. The methods adopted for this purpose include the modular construction of codes, simplified geometries, automatic variance reduction techniques, continuous energy treatment of cross section data, and albedo methods for streaming. Descriptions are given of the implementation of these methods and of their use in practical calculations. 26 references. (U.S.)

  1. Calculational approach to ionization spectrometer design

    International Nuclear Information System (INIS)

    Gabriel, T.A.

    1974-01-01

    Many factors contribute to the design and overall performance of an ionization spectrometer. These factors include the conditions under which the spectrometer is to be used, the required performance, the development of the hadronic and electromagnetic cascades, leakage and binding energies, saturation effects of densely ionizing particles, nonuniform light collection, sampling fluctuations, etc. The calculational procedures developed at Oak Ridge National Laboratory that have been applied to many spectrometer designs and that include many of the influencing factors in spectrometer design are discussed. The incident-particle types which can be considered with some generality are protons, neutrons, pions, muons, electrons, positrons, and gamma rays. Charged kaons can also be considered but with less generality. The incident-particle energy range can extend into the hundreds of GeV range. The calculations have been verified by comparison with experimental data but only up to approximately 30 GeV. Some comparisons with experimental data are also discussed and presented so that the flexibility of the calculational methods can be demonstrated. (U.S.)

  2. SMART core preliminary nuclear design-II

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

    1997-06-01

    Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

  3. Preliminary design of a tandem mirror reactor

    International Nuclear Information System (INIS)

    Strohmayer, J.N.

    1984-04-01

    The purpose of this thesis is to examine the TARA mirror experiment as a possible tandem mirror reactor configuration. This is a preliminary study to size the coil structure based on using the smallest end cell axial length that physics and engineering allow, zeroing the central cell parallel currents and having interchange stability. The input powers are estimated for the final reactor design so a Q value may be estimated. The Q value is defined as the fusion power divided by the total injected power absorbed by the plasma. A computer study was performed on the effect of the transition size, the transition vertical spacing and transition current. These parameters affect the central cell parallel currents, the recircularization of the flux tube and the ratio of central cell beta to anchor beta needed for marginal stability. Two designs were identified. The first uses 100 keV and 13 keV neutral beams to pump the ions that trap in the thermal barrier. The Q value of this reactor is 11.3. The second reactor uses a pump beam at 40 keV. This energy is chosen because there is a resonance for the charge exchange cross section between D 0 and He 2+ at this energy, thus the alpha ash will be pumped along with the deuterium and tritium. The Q value of this reactor is 11.6

  4. Preliminary design of the repository, stage 2

    International Nuclear Information System (INIS)

    Saanio, T.; Kirkkomaeki, T.; Keto, P.; Kukkola, T.; Raiko, H.

    2007-01-01

    Spent nuclear fuel from Finnish nuclear power plants will be disposed of in deep bedrock in Olkiluoto, Eurajoki. The repository is planned to be excavated at a depth of 400 - 500 metres. Access routes to the repository include a 1:10 inclined access tunnel, and vertical shafts. The fuel is encapsulated in the encapsulation plant above ground and transferred to the repository in the canister lift. Deposition tunnels, central tunnels and technical rooms are excavated at the disposal level. The canisters are deposited in deposition holes that are covered with bentonite blocks. The deposition holes are bored in the floors of the deposition tunnels. The central tunnel system consists of two parallel central tunnels that are inter-connected at certain distances. Two parallel central tunnels improve the fire safety of the rooms and also allow flexible backfilling and closing of the deposition tunnels in stages at the operational phase of the repository. An underground rock characterization facility, ONKALO, is excavated at the disposal level to support and confirm investigations carried out from above ground. ONKALO is designed so that it can later serve as part of the repository. ONKALO excavations were started in 2004. The repository will be excavated in the 2010s and operation will start in 2020. The fifth nuclear power unit makes the operational phase of the repository very long. Parts of the repository will be excavated and closed over the long operational period. The repository can be constructed at one or several levels. The one-storey alternative is the so-called reference alternative in this preliminary design report. The two-storey alternative is also taken into account in the ONKALO designs. The preliminary designs of the repository are presented as located in Olkiluoto. The location of the repository will be revised when more information on the bedrock has been gained. More detailed data of the circumstances will be obtained from above ground investigations

  5. Preliminary design of the repository. Stage 2

    International Nuclear Information System (INIS)

    Saanio, T.; Kirkkomaeki, T.; Keto, P.; Kukkola, T.; Raiko, H.

    2007-04-01

    Spent nuclear fuel from Finnish nuclear power plants will be disposed of in deep bedrock in Olkiluoto, Eurajoki. The repository is planned to be excavated at a depth of 400 - 500 metres. Access routes to the repository include a 1:10 inclined access tunnel, and vertical shafts. The fuel is encapsulated in the encapsulation plant above ground and transferred to the repository in the canister lift. Deposition tunnels, central tunnels and technical rooms are excavated at the disposal level. The canisters are deposited in deposition holes that are covered with bentonite blocks. The deposition holes are bored in the floors of the deposition tunnels. The central tunnel system consists of two parallel central tunnels that are inter-connected at certain distances. Two parallel central tunnels improve the fire safety of the rooms and also allow flexible backfilling and closing of the deposition tunnels in stages at the operational phase of the repository. An underground rock characterization facility, ONKALO, is excavated at the disposal level to support and confirm investigations carried out from above ground. ONKALO is designed so that it can later serve as part of the repository. ONKALO excavations were started in 2004. The repository will be excavated in the 2010s and operation will start in 2020. The fifth nuclear power unit makes the operational phase of the repository very long. Parts of the repository will be excavated and closed over the long operational period. The repository can be constructed at one or several levels. The one-storey alternative is the so-called reference alternative in this preliminary design report. The two-storey alternative is also taken into account in the ONKALO designs. The preliminary designs of the repository are presented as located in Olkiluoto. The location of the repository will be revised when more information on the bedrock has been gained. More detailed data of the circumstances will be obtained from above ground investigations

  6. Preliminary design report for the NAC combined transport cask

    International Nuclear Information System (INIS)

    1990-04-01

    Nuclear Assurance Corporation (NAC) is under contract to the United States Department of Energy (DOE) to design, license, develop and test models, and fabricate a prototype cask transportation system for nuclear spent fuel. The design of this combined transport (rail/barge) transportation system has been divided into two phases, a preliminary design phase and a final design phase. This Preliminary Design Package (PDP) describes the NAC Combined Transport Cask (NAC-CTC), the results of work completed during the preliminary design phase and identifies the additional detailed analyses, which will be performed during final design. Preliminary analytical results are presented in the appropriate sections and supplemented by summaries of procedures and assumptions for performing the additional detailed analyses of the final design. 60 refs., 1 fig., 2 tabs

  7. Methodology for Preliminary Design of Electrical Microgrids

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Richard P. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stamp, Jason E. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Eddy, John P. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Henry, Jordan M [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Munoz-Ramos, Karina [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Abdallah, Tarek [U.S. Army Corps of Engineers, Washington, DC (United States)

    2015-09-30

    Many critical loads rely on simple backup generation to provide electricity in the event of a power outage. An Energy Surety Microgrid TM can protect against outages caused by single generator failures to improve reliability. An ESM will also provide a host of other benefits, including integration of renewable energy, fuel optimization, and maximizing the value of energy storage. The ESM concept includes a categorization for microgrid value proposi- tions, and quantifies how the investment can be justified during either grid-connected or utility outage conditions. In contrast with many approaches, the ESM approach explic- itly sets requirements based on unlikely extreme conditions, including the need to protect against determined cyber adversaries. During the United States (US) Department of Defense (DOD)/Department of Energy (DOE) Smart Power Infrastructure Demonstration for Energy Reliability and Security (SPIDERS) effort, the ESM methodology was successfully used to develop the preliminary designs, which direct supported the contracting, construction, and testing for three military bases. Acknowledgements Sandia National Laboratories and the SPIDERS technical team would like to acknowledge the following for help in the project: * Mike Hightower, who has been the key driving force for Energy Surety Microgrids * Juan Torres and Abbas Akhil, who developed the concept of microgrids for military installations * Merrill Smith, U.S. Department of Energy SPIDERS Program Manager * Ross Roley and Rich Trundy from U.S. Pacific Command * Bill Waugaman and Bill Beary from U.S. Northern Command * Melanie Johnson and Harold Sanborn of the U.S. Army Corps of Engineers Construc- tion Engineering Research Laboratory * Experts from the National Renewable Energy Laboratory, Idaho National Laboratory, Oak Ridge National Laboratory, and Pacific Northwest National Laboratory

  8. Preliminary seismic design of dynamically coupled structural systems

    International Nuclear Information System (INIS)

    Pal, N.; Dalcher, A.W.; Gluck, R.

    1977-01-01

    In this paper, the analysis criteria for coupling and decoupling, which are most commonly used in nuclear design practice, are briefly reviewed and a procedure outlined and demonstrated with examples. Next, a criterion judged to be practical for preliminary seismic design purposes is defined. Subsequently, a technique compatible with this criterion is suggested. A few examples are presented to test the proposed procedure for preliminary seismic design purposes. Limitations of the procedure are also discussed and finally, the more important conclusions are summarized

  9. Preliminary design review: Brayton Isotope Power System

    International Nuclear Information System (INIS)

    The design aspects covered include flight system design, design criteria/margins/reliability, GDS design, system analysis, materials, system assembly procedure, and government furnished equipment-BTPS

  10. Preliminary radiation shielding design for BOOMERANG

    International Nuclear Information System (INIS)

    Donahue, Richard J.

    2002-01-01

    Preliminary radiation shielding specifications are presented here for the 3 GeV BOOMERANG Australian synchrotron light source project. At this time the bulk shield walls for the storage ring and injection system (100 MeV Linac and 3 GeV Booster) are considered for siting purposes

  11. Electromagnetic design calculation of the control rod drive mechanism

    International Nuclear Information System (INIS)

    Zhu Qirong; Zhu Jingchang

    1991-01-01

    Electromagnetic design calculation of the step-by-step magnetic jacking control rod drive mechanism includes magnetic field force calculation and design calculation of magnetomotive force for three electromagnetic iron and their coilds. The basic principle and method of electromagnetic design calculation had been expounded to take the lift magnet and lift coil for example

  12. East Area Irradiation Test Facility: Preliminary FLUKA calculations

    CERN Document Server

    Lebbos, E; Calviani, M; Gatignon, L; Glaser, M; Moll, M; CERN. Geneva. ATS Department

    2011-01-01

    In the framework of the Radiation to Electronics (R2E) mitigation project, the testing of electronic equipment in a radiation field similar to the one occurring in the LHC tunnel and shielded areas to study its sensitivity to single even upsets (SEU) is one of the main topics. Adequate irradiation test facilities are therefore required, and one installation is under consideration in the framework of the PS East area renovation activity. FLUKA Monte Carlo calculations were performed in order to estimate the radiation field which could be obtained in a mixed field facility using the slowly extracted 24 GeV/c proton beam from the PS. The prompt ambient dose equivalent as well as the equivalent residual dose rate after operation was also studied and results of simulations are presented in this report.

  13. TITAN Legal Weight Truck cask preliminary design report

    International Nuclear Information System (INIS)

    1990-04-01

    The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of this document includes the LWT cask with fuel baskets, impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. The results of the tradeoff studies and evaluations that were performed during the preliminary design are presented in Appendix A to this report. 51 figs., 17 tabs

  14. Preliminary design package for solar heating and hot water system

    Science.gov (United States)

    1976-01-01

    Two prototype solar heating and hot water systems for use in single-family dwellings or commercial buildings were designed. Subsystems included are: collector, storage, transport, hot water, auxiliary energy, and government-furnished site data acquisition. The systems are designed for Yosemite, California, and Pueblo, Colorado. The necessary information to evaluate the preliminary design for these solar heating and hot water systems is presented. Included are a proposed instrumentation plan, a training program, hazard analysis, preliminary design drawings, and other information about the design of the system.

  15. RHEIN, Modular System for Reactor Design Calculation

    International Nuclear Information System (INIS)

    Reiche, Christian; Barz, Hansulrich; Kunzmann, Bernd; Seifert, Eberhard; Wand, Hartmut

    1990-01-01

    1 - Description of program or function: RHEIN is a modular reactor code system for neutron physics calculations. It consists of a small number of system codes for execution control, data management, and handling support, as well as of the physical calculation routines. The execution is controlled by input data containing mathematical and physical parameters and simple commands for routine calls and data manipulations. The calculation routines are in tune with one another and the system takes care of the data transfer between them. Cross-section libraries with self shielding parameters are added to the system. 2 - Method of solution: The calculation routines can be used for solving the following physics problems: - Calculation of cross-section sets for infinite mediums, taking into account chord length. - Zero-dimensional spectrum calculation in diffusion, P1, or B1 approximation. - One-dimensional calculation in diffusion, P1, or collision probability approximation. - Two-dimensional diffusion calculation. - Cell calculation by THERMOS. - Zone-wise homogenized group collapsing within zero, one, or two-dimensional models. - Normalization, summarizing, etc. - Output of cross-section sets to off systems Sn and Monte-Carlo calculations

  16. Preliminary bridge design navigation tool for novices

    OpenAIRE

    Boulanger, Sylvie

    1997-01-01

    The motivation of the thesis comes from the frustrations of young engineers confronted with real design problems. The inspiration of the thesis evolved from observations of bridge designers and analyses of bridge design competitions. Not only do designers adopt more than one strategy during design, they rarely perform a fixed sequence of tasks. Not only do designers consider more than one criterion during design, their priorities shift during the determination of parameters. The choice of tas...

  17. Preliminary bridge design navigation tool for novices

    OpenAIRE

    Boulanger, Sylvie; Hirt, Manfred A.

    2008-01-01

    The motivation of the thesis comes from the frustrations of young engineers confronted with real design problems. The inspiration of the thesis evolved from observations of bridge designers and analyses of bridge design competitions. Not only do designers adopt more than one strategy during design, they rarely perform a fixed sequence of tasks. Not only do designers consider more than one criterion during design, their priorities shift during the determination of parameters. The choice of tas...

  18. Preliminary evaluation of FY98 KALIMER shielding design

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Woon; Kang, Chang Mu; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    This report describes a preliminary evaluation of the shielding design of FY98 KALIMER. The KALIMER shielding design includes the Inner Fixed Shield of a stainless cylinder located inside the support barrel; the Radial PSDRS Shields which are three B{sub 4}C cylinders located outside the support barrel at core level; the Lower IHX shield of a cylindrical B{sub 4}C plate located above the flow guide; and Inner and Outer IHX shields of B{sub 4}C cylinders located inside and outside of the support barrel, respectively. The DORT3.1 two-dimensional transport code was used to evaluate the KALIMER shielding design. The reactor system was represented by four axial zones, each of which was modeled in the R-Z geometry. The KAFAX-F22 library was used in the analyses, which was generated from the JEF-2.2 of OECD/NEA files for LMR applications by KAERI. The performance of the KALIMER shielding design is compared against the shielding design criteria. The results indicate that the support barrel, upper grid plate, and other reactor structures meet the maximum neutron fluence and DPA limits established in the shielding design criteria. Activities of the air effluent in the PSDRS were also evaluated and are shown to satisfy the maximum permissible concentration (MPC) limits in 10 CFR Part 20. In the future, the validation of the DORT model by a detailed three dimensional calculation such as MCNP and the justification of the current shielding design limits are needed. (author). 13 refs., 23 figs., 31 tabs.

  19. Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2017-07-01

    The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

  20. Aberrations in preliminary design of ITER divertor impurity influx monitor

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@jaea.go.jp [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Ogawa, Hiroaki [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Katsunuma, Atsushi; Kitazawa, Daisuke [Core Technology Center, Nikon Corporation, Yokohama 244-8533 (Japan); Ohmori, Keisuke [Customized Products Business Unit, Nikon Corporation, Mito 310-0843 (Japan)

    2015-12-15

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • The spot diagrams were suppressed within the core of receiving fiber. • The aberration of DIM is suppressed in the preliminary design. - Abstract: Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

  1. Aberrations in preliminary design of ITER divertor impurity influx monitor

    International Nuclear Information System (INIS)

    Kitazawa, Sin-iti; Ogawa, Hiroaki; Katsunuma, Atsushi; Kitazawa, Daisuke; Ohmori, Keisuke

    2015-01-01

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • The spot diagrams were suppressed within the core of receiving fiber. • The aberration of DIM is suppressed in the preliminary design. - Abstract: Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

  2. MR imaging of prostate. Preliminary experience with calculated imaging in 28 cases

    International Nuclear Information System (INIS)

    Gevenois, P.A.; Van Regemorter, G.; Ghysels, M.; Delepaut, A.; Van Gansbeke, D.; Struyven, J.

    1988-01-01

    The majority of studies with MR imaging in prostate disease are based on a semiology obtained using images weighted in T1 and T2. A study was carried out to evaluate effects of images calculated in T1 and T2 obtained at 0.5T. This preliminary study concerns 28 prostate examinations with spin-echo acquisition and inversion-recuperation parameters, and provided images calculated in T1, weighted and calculated in T2. Images allowed detection and characterization of prostate lesions. However, although calculated images accentuate discrimination of the method, the weighted images conserve their place because of their improved spatial resolution [fr

  3. Preliminary design package for prototype solar heating system

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    A summary is given of the preliminary analysis and design activity on solar heating systems. The analysis was made without site specific ata other than weather; therefore, the results indicate performance expected under these special conditions. Major items in this report include systeem candidates, design approaches, trade studies and other special data required to evaluate the preliminary analysis and design. The program calls for the development and delivery of eight prototype solar heating and coolin systems for installation and operational test. Two-heating and six heating and cooling units will be delivered for Single Family Residences (SFR), Multi-Family Residences (MFR) and commercial applications.

  4. Practical Recommendations for the Preliminary Design Analysis of ...

    African Journals Online (AJOL)

    Interior-to-exterior shear ratios for equal and unequal bay frames, as well as column inflection points were obtained to serve as practical aids for preliminary analysis/design of fixed-feet multistory sway frames. Equal and unequal bay five story frames were analysed to show the validity of the recommended design ...

  5. Preliminary design package for solar collector and solar pump

    Science.gov (United States)

    1978-01-01

    A solar-operated pump using an existing solar collector, for use on solar heating and cooling and hot water systems is described. Preliminary design criteria of the collector and solar-powered pump is given including: design drawings, verification plans, and hazard analysis.

  6. A preliminary study on the relevancy of sustainable building design ...

    African Journals Online (AJOL)

    This preliminary study aims to explore the relationship between sustainable building design paradigms and commercial property depreciation, to assist in the understanding of sustainable building design impact towards commercial building value and rental de employs the qualitative method and analyses valuers' current ...

  7. Preliminary System Design of the SWRL Financial System.

    Science.gov (United States)

    Ikeda, Masumi

    The preliminary system design of the computer-based Southwest Regional Laboratory's (SWRL) Financial System is outlined. The system is designed to produce various management and accounting reports needed to maintain control of SWRL operational and financial activities. Included in the document are descriptions of the various types of system…

  8. WIPP conceptual design report. Addendum A. Design calculations for Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    1977-04-01

    The design calculations for the Waste Isolation Pilot Plant (WIPP) are presented. The following categories are discussed: general nuclear calculations; radwaste calculations; structural calculations; mechanical calculations; civil calculations; electrical calculations; TRU waste surface facility time and motion analysis; shaft sinking procedures; hoist time and motion studies; mining system analysis; mine ventilation calculations; mine structural analysis; and miscellaneous underground calculations

  9. Calculations enable optimum design of magnetic brake

    Science.gov (United States)

    Kosmahl, H. G.

    1966-01-01

    Mathematical analysis and computations determine optimum magnetic coil configurations for a magnetic brake which controllably decelerates a free falling load to a soft stop. Calculations on unconventionally wound coils determine the required parameters for the desired deceleration with minimum electrical energy supplied to the stationary coil.

  10. Preliminary ALARA design concept for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyo Youn; Kim, Seung Nam; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    SMART(System-integrated Modular Advanced ReacTor) is a space saving integral type nuclear rector with the thermal power of 330 MW. This report provides general design guide and authority in NSSS designs for SMART needed to maintain the occupational doses and doses to members of public ALARA to meet the regulatory requirements. Paragraph 20.1 of 10 CFR 20, ''Standards for Protection Against Radiation'', states that licensee should make every reasonable effort to maintain exposures to radiation as far below the limits specified in Part 20 as is reasonably achievable. The ALARA (as low as is reasonably achievable) principle is incorporated into Korean radiation protection law as paragraph one Article 97 of the Atomic Energy Act. (Jan. 1995). This ALARA Design Concept for SMART provides 1) description of the organization and responsibilities needed for upper level management support and authority in order for the implementation of ALARA, 2) guidance and procedures for design, review, and evaluation needed for SMART ALARA program implementation, 3) general design guidelines for SMART NSSS and BOP designers to implement ALARA principles in design stage, and 4) training and instruction requirement of SMART NSSS and BOP designers for the familiarization of ALARA principles to be implemented in NSSS designs. (Author). 4 refs., 1 tabs.

  11. Preliminary ALARA design concept for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyo Youn; Kim, Seung Nam; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    SMART(System-integrated Modular Advanced ReacTor) is a space saving integral type nuclear rector with the thermal power of 330 MW. This report provides general design guide and authority in NSSS designs for SMART needed to maintain the occupational doses and doses to members of public ALARA to meet the regulatory requirements. Paragraph 20.1 of 10 CFR 20, ''Standards for Protection Against Radiation'', states that licensee should make every reasonable effort to maintain exposures to radiation as far below the limits specified in Part 20 as is reasonably achievable. The ALARA (as low as is reasonably achievable) principle is incorporated into Korean radiation protection law as paragraph one Article 97 of the Atomic Energy Act. (Jan. 1995). This ALARA Design Concept for SMART provides 1) description of the organization and responsibilities needed for upper level management support and authority in order for the implementation of ALARA, 2) guidance and procedures for design, review, and evaluation needed for SMART ALARA program implementation, 3) general design guidelines for SMART NSSS and BOP designers to implement ALARA principles in design stage, and 4) training and instruction requirement of SMART NSSS and BOP designers for the familiarization of ALARA principles to be implemented in NSSS designs. (Author). 4 refs., 1 tabs.

  12. Preliminary ALARA design concept for SMART

    International Nuclear Information System (INIS)

    Kim, Kyo Youn; Kim, Seung Nam; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    SMART(System-integrated Modular Advanced ReacTor) is a space saving integral type nuclear rector with the thermal power of 330 MW. This report provides general design guide and authority in NSSS designs for SMART needed to maintain the occupational doses and doses to members of public ALARA to meet the regulatory requirements. Paragraph 20.1 of 10 CFR 20, ''Standards for Protection Against Radiation'', states that licensee should make every reasonable effort to maintain exposures to radiation as far below the limits specified in Part 20 as is reasonably achievable. The ALARA (as low as is reasonably achievable) principle is incorporated into Korean radiation protection law as paragraph one Article 97 of the Atomic Energy Act. (Jan. 1995). This ALARA Design Concept for SMART provides 1) description of the organization and responsibilities needed for upper level management support and authority in order for the implementation of ALARA, 2) guidance and procedures for design, review, and evaluation needed for SMART ALARA program implementation, 3) general design guidelines for SMART NSSS and BOP designers to implement ALARA principles in design stage, and 4) training and instruction requirement of SMART NSSS and BOP designers for the familiarization of ALARA principles to be implemented in NSSS designs. (Author). 4 refs., 1 tabs

  13. Gas turbine designer computer program - a study of using a computer for preliminary design of gas turbines

    Energy Technology Data Exchange (ETDEWEB)

    Petersson, Rickard

    1995-11-01

    This thesis presents calculation schemes and theories for preliminary design of the fan, high pressure compressor and turbine of a gas turbine. The calculations are presented step by step, making it easier to implement in other applications. The calculation schemes have been implemented as a subroutine in a thermodynamic program. The combination of the thermodynamic cycle calculation and the design calculation turned out to give quite relevant results, when predicting the geometry and performance of an existing aero engine. The program developed is able to handle several different gas turbines, including those in which the flow is split (i.e. turbofan engines). The design process is limited to the fan, compressor and turbine of the gas turbine, the rest of the components have not been considered. Output from the program are main geometry, presented both numerically and as a scale plot, component efficiencies, stresses in critical points and a simple prediction of turbine blade temperatures. 11 refs, 21 figs, 1 tab

  14. Preliminary integrated calculation of radionuclide cation and anion transport at Yucca Mountain using a geochemical model

    International Nuclear Information System (INIS)

    Birdsell, K.H.; Campbell, K.; Eggert, K.G.; Travis, B.J.

    1989-01-01

    This paper presents preliminary transport calculations for radionuclide movement at Yucca Mountain using preliminary data for mineral distributions, retardation parameter distributions, and hypothetical recharge scenarios. These calculations are not performance assessments, but are used to study the effectiveness of the geochemical barriers at the site at mechanistic level. The preliminary calculations presented have many shortcomings and should be viewed only as a demonstration of the modeling methodology. The simulations were run with TRACRN, a finite-difference porous flow and radionuclide transport code developed for the Yucca Mountain Project. Approximately 30,000 finite-difference nodes are used to represent the unsaturated and saturated zones underlying the repository in three dimensions. Sorption ratios for the radionuclides modeled are assumed to be functions of mineralogic assemblages of the underlying rock. These transport calculations present a representative radionuclide cation, 135 Cs and anion, 99 Tc. The effects on transport of many of the processes thought to be active at Yucca Mountain may be examined using this approach. The model provides a method for examining the integration of flow scenarios, transport, and retardation processes as currently understood for the site. It will also form the basis for estimates of the sensitivity of transport calculations to retardation processes. 11 refs., 17 figs., 1 tab

  15. Shielding design at Fermilab: Calculations and measurements

    International Nuclear Information System (INIS)

    Cossairt, J.D.

    1986-11-01

    The development of the Fermilab accelerator complex during the past two decades from its concept as the ''200 BeV accelerator'' to that of the present tevatron, designed to operate at energies as high as 1 TeV, has required a coincidental refinement and development in methods of shielding design. In this paper I describe these methods as used by the radiation protection staff of Fermilab. This description will review experimental measurements which substantiate these techniques in realistic situations. Along the way, observations will be stated which likely are applicable to other protron accelerators in the multi-hundred GeV energy region, including larger ones yet to be constructed

  16. Preliminary A ampersand PCT multiple detector design

    International Nuclear Information System (INIS)

    Roberson, G.P.; Martz, H.E.; Camp, D.C.; Decman, D.J.; Johansson, E.M.

    1997-01-01

    The next generation, multi-detector active and passive computed tomography (A ampersand PCT) scanner will be optimized for speed and accuracy. At the Lawrence Livermore National Lab (LLNL) we have demonstrated the trade-offs between different A ampersand PCT design parameters that affect the speed and quality of the assay results. These fundamental parameters govern the optimum system design. Although the multi-detector scanner design has priority put on speed to increase waste drum throughput, higher speed should not compromise assay accuracy. One way to increase the speed of the A ampersand PCT technology is to use multiple detectors. This yields a linear speedup by a factor approximately equal to the number of detectors used without a compromise in system accuracy. There are many different design scenarios that can be developed using multiple detectors. Here we describe four different scenarios and discuss the trade-offs between them. Also, some considerations are given in this design description for the implementation of a multiple detector technology in a field- deployable mobile trailer system

  17. Computing tools for accelerator design calculations

    International Nuclear Information System (INIS)

    Fischler, M.; Nash, T.

    1984-01-01

    This note is intended as a brief, summary guide for accelerator designers to the new generation of commercial and special processors that allow great increases in computing cost effectiveness. New thinking is required to take best advantage of these computing opportunities, in particular, when moving from analytical approaches to tracking simulations. In this paper, we outline the relevant considerations

  18. A knowledge-based design framework for airplane conceptual and preliminary design

    Science.gov (United States)

    Anemaat, Wilhelmus A. J.

    The goal of work described herein is to develop the second generation of Advanced Aircraft Analysis (AAA) into an object-oriented structure which can be used in different environments. One such environment is the third generation of AAA with its own user interface, the other environment with the same AAA methods (i.e. the knowledge) is the AAA-AML program. AAA-AML automates the initial airplane design process using current AAA methods in combination with AMRaven methodologies for dependency tracking and knowledge management, using the TechnoSoft Adaptive Modeling Language (AML). This will lead to the following benefits: (1) Reduced design time: computer aided design methods can reduce design and development time and replace tedious hand calculations. (2) Better product through improved design: more alternative designs can be evaluated in the same time span, which can lead to improved quality. (3) Reduced design cost: due to less training and less calculation errors substantial savings in design time and related cost can be obtained. (4) Improved Efficiency: the design engineer can avoid technically correct but irrelevant calculations on incomplete or out of sync information, particularly if the process enables robust geometry earlier. Although numerous advancements in knowledge based design have been developed for detailed design, currently no such integrated knowledge based conceptual and preliminary airplane design system exists. The third generation AAA methods are tested over a ten year period on many different airplane designs. Using AAA methods will demonstrate significant time savings. The AAA-AML system will be exercised and tested using 27 existing airplanes ranging from single engine propeller, business jets, airliners, UAV's to fighters. Data for the varied sizing methods will be compared with AAA results, to validate these methods. One new design, a Light Sport Aircraft (LSA), will be developed as an exercise to use the tool for designing a new airplane

  19. Preliminary design of an asteroid hopping mission

    Science.gov (United States)

    Scheppa, Michael D.

    In 2010, NASA announced that its new vision is to support private space launch operations. It is anticipated that this new direction will create the need for new and innovative ideas that push the current boundaries of space exploration and contain the promise of substantial gain, both in research and capital. The purpose of the study is to plan and estimate the feasibility of a mission to visit a number of near Earth asteroids (NEAs). The mission would take place before the end of the 21st century, and would only use commercially available technology. Throughout the mission design process, while holding astronaut safety paramount, it was the goal to maximize the return while keeping the cost to a minimum. A mission of the nature would appeal to the private space industry because it could be easily adapted and set into motion. The mission design was divided into three main parts; mission timeline, vehicle design and power sources, with emphasis on nuclear and solar electric power, were investigated. The timeline and associated trajectories were initially selected using a numerical estimation and then optimized using Satellite Tool Kit (STK) 9.s's Design Explorer Optimizer [1]. Next, the spacecraft was design using commercially available parts that would support the mission requirements. The Variable Specific Impulse Magnetoplasma Rocket (VASIMR) was and instrumental piece in maximizing the number of NEAs visited. Once the spacecraft was designed, acceptable power supply options were investigated. The VASIMR VX-200 requires 200 kilowatts of power to maintain thrust. This creates the need for a substantial power supply that consists of either a nuclear reactor of massive solar arrays. STK 9.1's Design Explorer Optimizer was able to create a mission time line that allowed for the exploration of seven NEAs in under two years, while keeping the total mission DeltaV under 71 kilometers per second. Based on these initial findings, it is determined that a mission of this

  20. Design calculations for NIF convergent ablator experiments

    Directory of Open Access Journals (Sweden)

    Olson R.E.

    2013-11-01

    Full Text Available The NIF convergent ablation tuning effort is underway. In the early experiments, we have discovered that the design code simulations over-predict the capsule implosion velocity and shock flash ρr, but under-predict the hohlraum x-ray flux measurements. The apparent inconsistency between the x-ray flux and radiography data implies that there are important unexplained aspects of the hohlraum and/or capsule behavior.

  1. Preliminary design report for the prototypical fuel rod consolidation system

    International Nuclear Information System (INIS)

    Rosa, J.M.

    1986-01-01

    This report documents NUTECH's preliminary design of a dry, spent fuel rod consolidation system. This preliminary design is the result of Phase I of a planned four phase project. The present report on this project provides a considerable amount of detail for a preliminary design effort. The design and all of its details are described in this Preliminary Design Report (PDR). The NUTECH dry rod consolidation system described herein is remotely operated. It provides for automatic operation, but with operator hold points between key steps in the process. The operator has the ability to switch to a manual operation mode at any point in the process. The system is directed by the operator using an executive computer which controls and coordinates the operation of the in-cell equipment. The operator monitors the process using an in-cell closed circuit television (CCTV) system with audio output and equipment status displays on the computer monitor. The in-cell mechanical equipment consists of the following: (1) two overhead cranes with manipulators; (2) a multi-degree of freedom fuel handling table and its clamping equipment; (3) a fuel assembly end fitting removal station and its tools; (4) a consolidator (which pulls rods, assembles the consolidated bundle and loads the canister); (5) a canister end cap welder and weld inspection system; (6) decontamination systems; and (7) the CCTV and microphone systems

  2. Sewage Solids Irradiator Transportation System (SSITS) cask: preliminary design description

    International Nuclear Information System (INIS)

    Eakes, R.G.; Kempka, S.N.; Lamoreaux, G.H.; Sutherland, S.H.

    1983-02-01

    The preliminary design of the Sewage Solids Irradiator Transportation System (SSITS) Cask is presented in this document. The SSITS cask is to be used for the transport of radioactive cesium chloride and strontium fluoride capsules which are of use in irradiators or as heat sources. The SSITS cask is approximately 1.4 m in diameter, 1.3 m high, weighs roughly 9 t, provides 33 cm of steel shielding, and can dissipate up to 5.2 kW of decay heat. The cask design criteria are identified and a description of the cask design and operation is provided. Detailed analyses of the design were performed to demonstrate licensability of the cask by the Nuclear Regulatory Commission (NRC). Results of the analyses indicate that the preliminary design is in compliance with the pertinent regulatory requirements for licensing of a radioactive material transportation container

  3. Preliminary Design of the AEGIS Test Facility

    CERN Document Server

    Dassa, Luca; Cambiaghi, Danilo

    2010-01-01

    The AEGIS experiment is expected to be installed at the CERN Antiproton Decelerator in a very close future, since the main goal of the AEGIS experiment is the measurement of gravity impact on antihydrogen, which will be produced on the purpose. Antihydrogen production implies very challenging environmental conditions: at the heart of the AEGIS facility 50 mK temperature, 1e-12 mbar pressure and a 1 T magnetic field are required. Interfacing extreme cryogenics with ultra high vacuum will affect very strongly the design of the whole facility, requiring a very careful mechanical design. This paper presents an overview of the actual design of the AEGIS experimental facility, paying special care to mechanical aspects. Each subsystem of the facility – ranging from the positron source to the recombination region and the measurement region – will be shortly described. The ultra cold region, which is the most critical with respect to the antihydrogen formation, will be dealt in detail. The assembly procedures will...

  4. Preliminary Tritium Management Design Activities at ORNL

    International Nuclear Information System (INIS)

    Harrison, Thomas J.; Felde, David K.; Logsdon, Randall J.; McFarlane, Joanna; Qualls, A. L.

    2016-01-01

    Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritium mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year's efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.

  5. Preliminary Tritium Management Design Activities at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Logsdon, Randall J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McFarlane, Joanna [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritium mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.

  6. Preliminary SP-100/Stirling heat exchanger designs

    International Nuclear Information System (INIS)

    Schmitz, P.; Tower, L.; Blue, B.; Dunn, P.

    1994-01-01

    Analytic modeling of several heat exchanger concepts to couple the SP-100 nuclear reactor lithium loop and the Space Stirling Power Convertor (SSPC) was performed. Four 25 kWe SSPC's are used to produce the required 100 kW of electrical power. This design work focused on the interface between a single SSPC and the primary lithium loop. Manifolding to separate and collect the four channel flow was not modeled. This work modeled two separate types of heat exchanger interfaces (conductive coupling and radiative coupling) to explore their relative advantages and disadvantages. The minimum mass design of the conductively coupled concepts was 18 kg or 0.73 kg/kWe for a single 25 kWe convertor. The minimum mass radiatively coupled concept was 41 kg or 1.64 kg/kWe. The direct conduction heat exchanger provides a lighter weight system because of its ability to operate the Stirling convertor evaporator at higher heat fluxes than those attainable by the radiatively coupled systems. Additionally the conductively coupled concepts had relatively small volumes and provide potentially simpler assembly. Their disadvantages were the tight tolerances and material joining problems associated with this refractory to superalloy interface. The advantages of the radiatively coupled designs were the minimal material interface problems

  7. Preliminary systems design study assessment report

    International Nuclear Information System (INIS)

    Mayberry, J.L.; Feizollahi, F.; Del Signore, J.C.

    1991-09-01

    The System Design Study (SDS), part of the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examined techniques available for the remediation of hazardous and transuranic waste stored at the Radioactive Waste Management Complex's Subsurface Disposal Area at the INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. The SDS resulted in the development of technology requirements including demonstration, testing, and evaluation activities needed for implementing each concept

  8. Preliminary design review report for K Basin Dose Reduction Project

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1996-01-01

    The strategy for reducing radiation dose, originating from radionuclides absorbed in the K East Basin concrete, is to raise the pool water level to provide additional shielding. This report documents a preliminary design review conducted to ensure that design approaches for cleaning/coating basin walls and modifying other basin components were appropriate. The conclusion of this review was that design documents presently conclusion of this review was that design documents presently completed or in process of modification are and acceptable basis for proceeding to complete the design

  9. TITAN Legal Weight Truck cask preliminary design report

    International Nuclear Information System (INIS)

    1990-04-01

    The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of the document includes the LWT cask with fuel baskets; impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. 75 figs., 48 tabs

  10. Preliminary design package for solar heating and hot water system

    Science.gov (United States)

    1977-01-01

    The preliminary design review on the development of a multi-family solar heating and domestic hot water prototype system is presented. The report contains the necessary information to evaluate the system. The system consists of the following subsystems: collector, storage, transport, control and Government-furnished site data acquisition.

  11. Preliminary Systems Design Study assessment report

    International Nuclear Information System (INIS)

    Mayberry, J.L.; Feizollahi, F.; Del Signore, J.C.

    1992-01-01

    The System Design Study (SDS), part of the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examined techniques available for the remediation of hazardous and transuranic waste stored at the Radioactive Waste Management Complex's Subsurface Disposal Area at the INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. The SDS resulted in the development of technology requirements including demonstration, testing, and evaluation activities needed for implementing each concept. This volume of the Systems Design Study contain four Appendixes that were part of the study. Appendix A is an EG ampersand G Idaho, Inc., report that represents a review and compilation of previous reports describing the wastes and quantities disposed in the Subsurface Disposal Area of the Idaho National Engineering Laboratory. Appendix B contains the process flowsheets considered in this study, but not selected for detailed analysis. Appendix C is a historical tabulation of radioactive waste incinerators. Appendix D lists Department of Energy facilities where cementation stabilization systems have been used

  12. Preliminary Mechanical Design of FHX for PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Jinyup; Koo, G. H.; Kim, S. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this paper, more specific data from analysis and mechanical method of approach to design will be addressed. Especially, frame of tube bundle and housing of FHX. Heretofore, it is concept design by mechanical basic knowledge and research of various structures that are activating in realities. Especially, to reduce thermal stress, we have planning to attach insulations inside the housing. In as much as FHX is as important on SFR as the other part, hereafter, we will develop FEM to check feasibility of the FHX's housing design in order to perform static and thermal analysis as well as bucking, seismic and so on. The Forced-draft sodium-to-air Heat Exchanger system (FHX) (employed in the Active Decay Heat Removal System (ADHRS) is a shell-and-tube type counter-current flow heat exchanger with serpentine finned-tube arrangement. Liquid sodium flows over the finned tubes. The unit is placed above the reactor building and has function of dumping the system heat load into the final heat sink, i. e., the atmosphere. Heat is transmitted from the primary hot sodium pool into the ADHRS sodium loop via Decay Heat Exchanger (DHX), and a direct heat exchange occurs between the tube-side sodium and the shell-side air through the FHX sodium tube wall. Cold atmospheric air is introduced into the air inlet duct at the lower part of the unit by using an electrically driven air blower. Air flows across the finned tube bank rising upward direction to make uniform air flow with perfect mixing across the tubes. The finned tube bundle is placed inside a well-insulated casing. The air heated at the tube bank region is collected at the top of the unit and then is discharged through the air stack above the unit. Although a blower supplies atmospheric air into the FHX unit, a tall air stack is also provided to secure natural draft head of natural circulation air flow against a loss power supply. The stack also has rain protecting structures to prevent inflow of rain drops or undesired

  13. Preliminary systems design study assessment report

    International Nuclear Information System (INIS)

    Mayberry, J.L.; Feizollahi, F.; Del Signore, J.C.

    1992-01-01

    The System Design Study (SDS), part of the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examined techniques available for the remediation of hazardous and transuranic waste stored at the Radioactive Waste Management Complex's Subsurface Disposal Area at the INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. The SDS resulted in the development of technology requirements including demonstration, testing, and evaluation activities needed for implementing each. This volume contains the descriptions and other relevant information of the four subsystems required for most of the ex situ processing systems. This volume covers the metal decontamination and sizing subsystem, soils processing subsystem, low-level waste subsystem, and retrieval subsystem

  14. Preliminary Systems Design Study assessment report

    International Nuclear Information System (INIS)

    Mayberry, J.L.; Quapp, W.J.; Feizollahi, F.; Del Signore, J.C.

    1991-07-01

    The System Design Study (SDS), part of the Waste Technology Development Department at Idaho National Engineering Laboratory (INEL), examined techniques available for the remediation of hazardous and transuranic (TRU) waste stored at the Radioactive Waste Management Complex's (RWMC's) Subsurface Disposal Area (SDA) at INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. SDS resulted in the development of technology requirements including demonstration, testing and evaluation activities needed for implementing each concept. The SDS results are published in eight volumes. Volume 1 contains an executive summary. The SDS summary and analysis of results are presented in volume 2. Volumes 3 through 7 contain detailed descriptions of twelve system and four subsystem concepts. Volume 8 contains the appendices. 3 figs., 3 tabs

  15. Preliminary Systems Design Study assessment report

    International Nuclear Information System (INIS)

    Mayberry, J.L.; Quapp, W.J.; Feizollahi, F.; Del Signore, J.C.

    1991-07-01

    The System Design Study (SDS), part of the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examined techniques available for the remediation of hazardous and transuranic waste stored at the Radioactive Waste Management Complex's Subsurface Disposal Area at the INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. SDS resulted in the development of technology requirements including demonstration, testing, and evaluation activities needed for implementing each concept. The SDS results are published in eight volumes. Volume 1 contains an executive summary. The SDS summary and analysis of results are presented in Volume 2. Volumes 3 through 7 contain detailed descriptions of twelve system and four subsystem concepts. Volume 8 contains the appendixes. 23 refs., 23 figs., 16 tabs

  16. Preliminary 2D design study for A ampersand PCT

    International Nuclear Information System (INIS)

    Keto, E.; Azevedo, S.; Roberson, P.

    1995-03-01

    Lawrence Livermore National Laboratory is currently designing and constructing a tomographic scanner to obtain the most accurate possible assays of radioactivity in barrels of nuclear waste in a limited amount of time. This study demonstrates a method to explore different designs using laboratory experiments and numerical simulations. In particular, we examine the trade-off between spatial resolution and signal-to-noise. The simulations are conducted in two dimensions as a preliminary study for three dimensional imaging. We find that the optimal design is entirely dependent on the expected source sizes and activities. For nuclear waste barrels, preliminary results indicate that collimators with widths of 1 to 3 inch and aspect ratios of 5:1 to 10:1 should perform well. This type of study will be repeated in 3D in more detail to optimize the final design

  17. Preliminary Design of Aerial Spraying System for Microlight Aircraft

    Science.gov (United States)

    Omar, Zamri; Idris, Nurfazliawati; Rahim, M. Zulafif

    2017-10-01

    Undoubtedly agricultural is an important sector because it provides essential nutrients for human, and consequently is among the biggest sector for economic growth worldwide. It is crucial to ensure crops production is protected from any plant diseases and pests. Thus aerial spraying system on crops is developed to facilitate farmers to for crops pests control and it is very effective spraying method especially for large and hilly crop areas. However, the use of large aircraft for aerial spaying has a relatively high operational cost. Therefore, microlight aircraft is proposed to be used for crops aerial spraying works for several good reasons. In this paper, a preliminary design of aerial spraying system for microlight aircraft is proposed. Engineering design methodology is adopted in the development of the aerial sprayer and steps involved design are discussed thoroughly. A preliminary design for the microlight to be attached with an aerial spraying system is proposed.

  18. Design Preliminaries for Direct Drive under Water Wind Turbine Generator

    DEFF Research Database (Denmark)

    Leban, Krisztina Monika; Ritchie, Ewen; Argeseanu, Alin

    2012-01-01

    This paper focuses on the preliminary design process of a 20 MW electric generator. The application calls for an offshore, vertical axis, direct drive wind turbine. Arguments for selecting the type of electric machine for the application are presented and discussed. Comparison criteria for deciding...... on a type of machine are listed. Additional constraints emerging from the direct drive, vertical axis concepts are considered. General rules and a preliminary algorithm are discussed for the machine selected to be most suitable for the imposed conditions....

  19. Preliminary Systems Design Study assessment report

    International Nuclear Information System (INIS)

    Mayberry, J.L.; Feizollahi, F.; Del Signore, J.C.

    1992-01-01

    The System Design Study (SDS), part of the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examined techniques for the remediation of hazardous and transuranic waste stored at Radioactive Waste Management Complex's Subsurface Disposal Area at the INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. The SDS resulted in the development of technology requirements including demonstration, testing, and evaluation activities needed for implementing each concept. This volume contains introduction section containing a brief SDS background and lists the general assumptions and considerations used during the development of the system concepts. The introduction section is followed by sections describing two system concepts that produce a waste form in compliance with the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC) and transportation package (TRAMPAC) requirements. This system concept category is referred to as Waste Form 4, ''WIPP and TRAMPAC Acceptable.'' The following two system concepts are under this category: Sort, Treat, and Repackage System (4-BE-2); Volume Reduction and Packaging System (4-BE-4)

  20. Preliminary design study for a corkscrew gantry

    International Nuclear Information System (INIS)

    Koehler, A.M.

    1987-01-01

    For two years or more a group including the author has been working together to study some problems related to the design of a gentry system for flexible direction of a proton beam for clinical treatments. Some consideration was given to the classic gantry geometry. Attempting to reduce the radius of the gantry arm by reducing the drift space after the scattering foils led to an analysis of the significance of inverse square intensity effects. The conclusion reached is that a drift space of about 3 meters is required to preserve some skin sparing for larger targets. To circumvent this problem the scattering foils ere put somewhere inside or even before the gantry system, accepting the fact that magnet apertures would have to be increased. This gantry system has the interesting ability to produce oblong fields of excellent uniformity with reasonable efficiency, preferentially with the long axis of the field parallel to the axis of rotation. It was disappointing, however, to find that the overall size of the gantry with its counterweights remained very large. Another change in geometry was proposed therefore in order to reduce the space taken up by the gantry and its counterweight. The beam is bent 45 0 in the horizontal plane and then again by 45 0 so that it is pointing away from isocenter, but in the plan of rotation of the gantry. The beam is now bent in that plane of rotation until it is pointed at isocenter. This is accomplished by two bends of 135 0 each with a suitable drift space between them so that the beam is pointed vertically downward at isocenter. The three dimensional complexity of the beam trajectory led to the name Corkscrew Gantry

  1. Verification of EPA's " Preliminary remediation goals for radionuclides" (PRG) electronic calculator

    Energy Technology Data Exchange (ETDEWEB)

    Stagich, B. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-29

    The U.S. Environmental Protection Agency (EPA) requested an external, independent verification study of their “Preliminary Remediation Goals for Radionuclides” (PRG) electronic calculator. The calculator provides information on establishing PRGs for radionuclides at Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) sites with radioactive contamination (Verification Study Charge, Background). These risk-based PRGs set concentration limits using carcinogenic toxicity values under specific exposure conditions (PRG User’s Guide, Section 1). The purpose of this verification study is to ascertain that the computer codes has no inherit numerical problems with obtaining solutions as well as to ensure that the equations are programmed correctly.

  2. The 'Reacteur Jules Horowitz': The preliminary design

    International Nuclear Information System (INIS)

    Ballagny, A.; Frachet, S.; Minguet, J.L.; Leydier, C.

    1999-01-01

    The 'Reactor Jules Horowitz' is a new research reactor project dedicated to materials and nuclear fuels testing, the location of which is foreseen at the CEA-Cadarache site, and the start-up in 2008. The launching of this project arises from a double finding: 1) the development of nuclear power plants aimed at satisfying the energy needs of the next century cannot be envisaged without the disposal of experimental reactors which are unrivalled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. 2) the present park of experimental reactors is 30 to 40 years old and it is advisable to examine henceforth the necessity and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a reflection on the mid and long term irradiations needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfil the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The selected reactor project, among several different concepts, is finally a light water open pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows many irradiations in the core and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the main items, and have permit to determine: the core and fuel designs, with added pressurisation; the different core

  3. Gemini Planet Imager: Preliminary Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Macintosh, B

    2007-05-10

    completely limited by quasi-static wave front errors, so that contrast does not improve with integration times longer than about 1 minute. Using the rotation of the Earth to distinguish companions from artifacts or multiwavelength imaging improves this somewhat, but GPI will still need to surpass the performance of existing systems by one to two orders of magnitude--an improvement comparable to the transition from photographic plates to CCDs. This may sound daunting, but other areas of optical science have achieved similar breakthroughs, for example, the transition to nanometer-quality optics for extreme ultraviolet lithography, the development of MEMS wave front control devices, and the ultra-high contrast demonstrated by JPL's High Contrast Imaging Test-bed. In astronomy, the Sloan Digital Sky Survey, long baseline radio interferometry, and multi-object spectrographs have led to improvements of similar or greater order of magnitude. GPI will be the first project to apply these revolutionary techniques to ground-based astronomy, with a systems engineering approach that studies the impact of every design decision on the key metric--final detectable planet contrast.

  4. Preliminary design study of a steady state tokamak device

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Nakajima, Shinji; Ushigusa, Kenkichi; and athors)

    1992-09-01

    Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

  5. A preliminary conceptual design study for Korean fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

    2013-10-15

    Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

  6. Fast reactor calculational route for Pu burning core design

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, S. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  7. Preliminary design study of the TMT Telescope structure system: overview

    Science.gov (United States)

    Usuda, Tomonori; Ezaki, Yutaka; Kawaguchi, Noboru; Nagae, Kazuhiro; Kato, Atsushi; Takaki, Junji; Hirano, Masaki; Hattori, Tomoya; Tabata, Masaki; Horiuchi, Yasushi; Saruta, Yusuke; Sofuku, Satoru; Itoh, Noboru; Oshima, Takeharu; Takanezawa, Takashi; Endo, Makoto; Inatani, Junji; Iye, Masanori; Sadjadpour, Amir; Sirota, Mark; Roberts, Scott; Stepp, Larry

    2014-07-01

    We present an overview of the preliminary design of the Telescope Structure System (STR) of Thirty Meter Telescope (TMT). NAOJ was given responsibility for the TMT STR in early 2012 and engaged Mitsubishi Electric Corporation (MELCO) to take over the preliminary design work. MELCO performed a comprehensive preliminary design study in 2012 and 2013 and the design successfully passed its Preliminary Design Review (PDR) in November 2013 and April 2014. Design optimizations were pursued to better meet the design requirements and improvements were made in the designs of many of the telescope subsystems as follows: 1. 6-legged Top End configuration to support secondary mirror (M2) in order to reduce deformation of the Top End and to keep the same 4% blockage of the full aperture as the previous STR design. 2. "Double Lower Tube" of the elevation (EL) structure to reduce the required stroke of the primary mirror (M1) actuators to compensate the primary mirror cell (M1 Cell) deformation caused during the EL angle change in accordance with the requirements. 3. M1 Segment Handling System (SHS) to be able to make removing and installing 10 Mirror Segment Assemblies per day safely and with ease over M1 area where access of personnel is extremely difficult. This requires semi-automatic sequence operation and a robotic Segment Lifting Fixture (SLF) designed based on the Compliance Control System, developed for controlling industrial robots, with a mechanism to enable precise control within the six degrees of freedom of position control. 4. CO2 snow cleaning system to clean M1 every few weeks that is similar to the mechanical system that has been used at Subaru Telescope. 5. Seismic isolation and restraint systems with respect to safety; the maximum acceleration allowed for M1, M2, tertiary mirror (M3), LGSF, and science instruments in 1,000 year return period earthquakes are defined in the requirements. The Seismic requirements apply to any EL angle, regardless of the

  8. Second preliminary design of JAERI experimental fusion reactor (JXFR)

    International Nuclear Information System (INIS)

    Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

    1979-06-01

    Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

  9. Development of an X-ray installation for the study of secondary electrons: preliminary measurements and calculations

    International Nuclear Information System (INIS)

    Baguena, A.; Shaw, M.; Williart, A.; Baguena, A.; Garcia, G.

    2006-01-01

    We describe the calculations and preliminary measures made for the installation of a X-ray generator tube. This device is going to be used for the secondary electron production from photonic primary radiation of up to 125 keV. With this experimental system, we will study the energetic and space distribution of produced secondary electrons by obtaining its spectrum of energies and its angular distribution. This method of measurement is going to be applied in different targets of radiological, environmental and biological interest. Calculations in the present article include: theoretical yield of X-rays production of the designed equipment, necessary shielding for the radiological safety of the installation staff, and an estimated dose due to their use. Characteristics of the installation and the equipment are described with this purpose. (author)

  10. Development of an X-ray installation for the study of secondary electrons: preliminary measurements and calculations

    Energy Technology Data Exchange (ETDEWEB)

    Baguena, A.; Shaw, M.; Williart, A. [Universidad Nacional de Educacion a Distancia, Dpto. Fisica de los Materiales, Madrid (Spain); Baguena, A. [Consejo de Seguridad Nuclear, Madrid (Spain); Garcia, G. [Instituto de Matematicas y Fisica Fundamental, Consejo Superior de Investigaciones Cientificas, Madrid (Spain)

    2006-07-01

    We describe the calculations and preliminary measures made for the installation of a X-ray generator tube. This device is going to be used for the secondary electron production from photonic primary radiation of up to 125 keV. With this experimental system, we will study the energetic and space distribution of produced secondary electrons by obtaining its spectrum of energies and its angular distribution. This method of measurement is going to be applied in different targets of radiological, environmental and biological interest. Calculations in the present article include: theoretical yield of X-rays production of the designed equipment, necessary shielding for the radiological safety of the installation staff, and an estimated dose due to their use. Characteristics of the installation and the equipment are described with this purpose. (author)

  11. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

    2002-01-01

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  12. The Mixed Waste Management Facility. Preliminary design review

    International Nuclear Information System (INIS)

    1995-01-01

    This document presents information about the Mixed Waste Management Facility. Topics discussed include: cost and schedule baseline for the completion of the project; evaluation of alternative options; transportation of radioactive wastes to the facility; capital risk associated with incineration; radioactive waste processing; scaling of the pilot-scale system; waste streams to be processed; molten salt oxidation; feed preparation; initial operation to demonstrate selected technologies; floorplans; baseline revisions; preliminary design baseline; cost reduction; and project mission and milestones

  13. OSU TOMF Program Site Selection and Preliminary Concept Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Spadling, Steve [Oklahoma State Univ., Stillwater, OK (United States)

    2012-05-10

    The purpose of this report is to confirm the programmatic requirements for the new facilities, identify the most appropriate project site, and develop preliminary site and building concepts that successfully address the overall project goals and site issues. These new facilities will be designed to accommodate the staff, drivers and maintenance requirements for the future mixed fleet of passenger vehicles, Transit Style Buses and School Buses.

  14. Preliminary design package for solar hot water system

    Energy Technology Data Exchange (ETDEWEB)

    Fogle, Val; Aspinwall, David B.

    1977-12-01

    The information necessary to evaluate the preliminary design of the Solar Engineering and Manufacturing Company's (SEMCO) solar hot water system is presented. This package includes technical information, schematics, drawings and brochures. This system, being developed by SEMCO, consists of the following subsystems: collector, storage, transport, control, auxiliary energy, and Government-furnished site data acquisition. The two units being manufactured will be installed at Loxahatchee, Florida, and Macon, Georgia.

  15. Recent improvements in the calculation of prompt fission neutron spectra: Preliminary results

    International Nuclear Information System (INIS)

    Madland, D.G.; LaBauve, R.J.; Nix, J.R.

    1989-01-01

    We consider three topics in the refinement and improvement of our original calculations of prompt fission neutron spectra. These are an improved calculation of the prompt fission neutron spectrum N(E) from the spontaneous fission of 252 Cf, a complete calculation of the prompt fission neutron spectrum matrix N(E,E n ) from the neutron-induced fission of 235 U, at incident neutron energies ranging from 0 to 15 MeV, and an assessment of the scission neutron component of the prompt fission neutron spectrum. Preliminary results will be presented and compared with experimental measurements and an evaluation. A suggestion is made for new integral cross section measurements. (author). 45 refs, 12 figs, 1 tab

  16. Deep Underground Science and Engineering Laboratory - Preliminary Design Report

    CERN Document Server

    Lesko, Kevin T; Alonso, Jose; Bauer, Paul; Chan, Yuen-Dat; Chinowsky, William; Dangermond, Steve; Detwiler, Jason A; De Vries, Syd; DiGennaro, Richard; Exter, Elizabeth; Fernandez, Felix B; Freer, Elizabeth L; Gilchriese, Murdock G D; Goldschmidt, Azriel; Grammann, Ben; Griffing, William; Harlan, Bill; Haxton, Wick C; Headley, Michael; Heise, Jaret; Hladysz, Zbigniew; Jacobs, Dianna; Johnson, Michael; Kadel, Richard; Kaufman, Robert; King, Greg; Lanou, Robert; Lemut, Alberto; Ligeti, Zoltan; Marks, Steve; Martin, Ryan D; Matthesen, John; Matthew, Brendan; Matthews, Warren; McConnell, Randall; McElroy, William; Meyer, Deborah; Norris, Margaret; Plate, David; Robinson, Kem E; Roggenthen, William; Salve, Rohit; Sayler, Ben; Scheetz, John; Tarpinian, Jim; Taylor, David; Vardiman, David; Wheeler, Ron; Willhite, Joshua; Yeck, James

    2011-01-01

    The DUSEL Project has produced the Preliminary Design of the Deep Underground Science and Engineering Laboratory (DUSEL) at the rehabilitated former Homestake mine in South Dakota. The Facility design calls for, on the surface, two new buildings - one a visitor and education center, the other an experiment assembly hall - and multiple repurposed existing buildings. To support underground research activities, the design includes two laboratory modules and additional spaces at a level 4,850 feet underground for physics, biology, engineering, and Earth science experiments. On the same level, the design includes a Department of Energy-shepherded Large Cavity supporting the Long Baseline Neutrino Experiment. At the 7,400-feet level, the design incorporates one laboratory module and additional spaces for physics and Earth science efforts. With input from some 25 science and engineering collaborations, the Project has designed critical experimental space and infrastructure needs, including space for a suite of multi...

  17. Preliminary design concepts for the advanced neutron source reactor systems

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1988-01-01

    This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor

  18. Calculation and design for SSRF's bulk shield

    Energy Technology Data Exchange (ETDEWEB)

    Fang, K.M. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)]. E-mail: fangkm@sinap.ac.cn; Xu, X.J. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China); Cai, J.H. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)

    2006-12-15

    Shielding design objectives for the SSRF are chosen, assumptions for beam loss rates are given, the methods used on the APS by Moe are summarized and introduced to make calculation and design on bulk shield, the factor of skyshine is also considered, design thicknesses for SSRF's bulk shield are presented.

  19. A Generative Computer Model for Preliminary Design of Mass Housing

    Directory of Open Access Journals (Sweden)

    Ahmet Emre DİNÇER

    2014-05-01

    Full Text Available Today, we live in what we call the “Information Age”, an age in which information technologies are constantly being renewed and developed. Out of this has emerged a new approach called “Computational Design” or “Digital Design”. In addition to significantly influencing all fields of engineering, this approach has come to play a similar role in all stages of the design process in the architectural field. In providing solutions for analytical problems in design such as cost estimate, circulation systems evaluation and environmental effects, which are similar to engineering problems, this approach is being used in the evaluation, representation and presentation of traditionally designed buildings. With developments in software and hardware technology, it has evolved as the studies based on design of architectural products and production implementations with digital tools used for preliminary design stages. This paper presents a digital model which may be used in the preliminary stage of mass housing design with Cellular Automata, one of generative design systems based on computational design approaches. This computational model, developed by scripts of 3Ds Max software, has been implemented on a site plan design of mass housing, floor plan organizations made by user preferences and facade designs. By using the developed computer model, many alternative housing types could be rapidly produced. The interactive design tool of this computational model allows the user to transfer dimensional and functional housing preferences by means of the interface prepared for model. The results of the study are discussed in the light of innovative architectural approaches.

  20. PRELIMINARY STRUCTURAL OPTIMIZATION OF SOME FUMONISIN METABOLITES BY DENSITY FUNCTIONAL THEORY CALCULATION

    Directory of Open Access Journals (Sweden)

    István Bors

    2015-09-01

    Full Text Available Maize (Zea mays L. is often contaminated with Fusarium verticillioides. This harmful fungus produces fumonisins as secondary metabolites. These fumonisins can appear both free and hidden form in planta. The hidden form is usually bound covalently to cereal starch. From the hidden fumonisins, during enzymatic degradation, glycosides are formed, and the fumonisin is further decomposed during a de-esterification step. In this short communication some preliminary DFT calculated structural results which could be useful in the future to help to understand the van der Waals force controlled molecular interactions between these kinds of mycotoxin molecules and enzymes are demonstrated.

  1. Preliminary thermal design of the COLD-SAT spacecraft

    Science.gov (United States)

    Arif, Hugh

    1991-01-01

    The COLD-SAT free-flying spacecraft was to perform experiments with LH2 in the cryogenic fluid management technologies of storage, supply and transfer in reduced gravity. The Phase A preliminary design of the Thermal Control Subsystem (TCS) for the spacecraft exterior and interior surfaces and components of the bus subsystems is described. The TCS was composed of passive elements which were augmented with heaters. Trade studies to minimize the parasitic heat leakage into the cryogen storage tanks are described. Selection procedure for the thermally optimum on-orbit spacecraft attitude was defined. TRASYS-2 and SINDA'85 verification analysis was performed on the design and the results are presented.

  2. AGC-1 Experiment and Final Preliminary Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Robert L. Bratton; Tim Burchell

    2006-08-01

    This report details the experimental plan and design as of the preliminary design review for the Advanced Test Reactor Graphite Creep-1 graphite compressive creep capsule. The capsule will contain five graphite grades that will be irradiated in the Advanced Test Reactor at the Idaho National Laboratory to determine the irradiation induced creep constants. Seven other grades of graphite will be irradiated to determine irradiated physical properties. The capsule will have an irradiation temperature of 900 C and a peak irradiation dose of 5.8 x 10{sup 21} n/cm{sup 2} [E > 0.1 MeV], or 4.2 displacements per atom.

  3. Aberration analysis calculations for synchrotron radiation beamline design

    International Nuclear Information System (INIS)

    McKinney, W.R.; Howells, M.; Padmore, H.A.

    1997-09-01

    The application of ray deviation calculations based on aberration coefficients for a single optical surface for the design of beamline optical systems is reviewed. A systematic development is presented which allows insight into which aberration may be causing the rays to deviate from perfect focus. A new development allowing analytical calculation of line shape is presented

  4. TPX: Contractor preliminary design review. Volume 3, Design and analysis

    International Nuclear Information System (INIS)

    1995-01-01

    Several models have been formed for investigating the maximum electromagnetic loading and magnetic field levels associated with the Tokamak Physics eXperiment (TPX) superconducting Poloidal Field (PF) coils. The analyses have been performed to support the design of the individual fourteen hoop coils forming the PF system. The coils have been sub-divided into three coil systems consisting of the central solenoid (CS), PF5 coils, and the larger radius PF6 and PF7 coils. Various electromagnetic analyses have been performed to determine the electromagnetic loadings that the coils will experience during normal operating conditions, plasma disruptions, and fault conditions. The loadings are presented as net body forces acting individual coils, spatial variations throughout the coil cross section, and force variations along the path of the conductor due to interactions with the TF coils. Three refined electromagnetic models of the PF coil system that include a turn-by-turn description of the fields and forces during a worst case event are presented in this report. A global model including both the TF and PF system was formed to obtain the force variations along the path of the PF conductors resulting from interactions with the TF currents. In addition to spatial variations, the loadings are further subdivided into time-varying and steady components so that structural fatigue issues can be addressed by designers and analysts. Other electromagnetic design issues such as the impact of the detailed coil designs on field errors are addressed in this report. Coil features that are analyzed include radial transitions via short jogs vs. spiral type windings and the effects of layer-to-layer rotations (i.e clocking) on the field errors

  5. TPX: Contractor preliminary design review. Volume 3, Design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-30

    Several models have been formed for investigating the maximum electromagnetic loading and magnetic field levels associated with the Tokamak Physics eXperiment (TPX) superconducting Poloidal Field (PF) coils. The analyses have been performed to support the design of the individual fourteen hoop coils forming the PF system. The coils have been sub-divided into three coil systems consisting of the central solenoid (CS), PF5 coils, and the larger radius PF6 and PF7 coils. Various electromagnetic analyses have been performed to determine the electromagnetic loadings that the coils will experience during normal operating conditions, plasma disruptions, and fault conditions. The loadings are presented as net body forces acting individual coils, spatial variations throughout the coil cross section, and force variations along the path of the conductor due to interactions with the TF coils. Three refined electromagnetic models of the PF coil system that include a turn-by-turn description of the fields and forces during a worst case event are presented in this report. A global model including both the TF and PF system was formed to obtain the force variations along the path of the PF conductors resulting from interactions with the TF currents. In addition to spatial variations, the loadings are further subdivided into time-varying and steady components so that structural fatigue issues can be addressed by designers and analysts. Other electromagnetic design issues such as the impact of the detailed coil designs on field errors are addressed in this report. Coil features that are analyzed include radial transitions via short jogs vs. spiral type windings and the effects of layer-to-layer rotations (i.e clocking) on the field errors.

  6. Preliminary design study of a large scale graphite oxidation loop

    International Nuclear Information System (INIS)

    Epel, L.G.; Majeski, S.J.; Schweitzer, D.G.; Sheehan, T.V.

    1979-08-01

    A preliminary design study of a large scale graphite oxidation loop was performed in order to assess feasibility and to estimate capital costs. The nominal design operates at 50 atmospheres helium and 1800 F with a graphite specimen 30 inches long and 10 inches in diameter. It was determined that a simple single walled design was not practical at this time because of a lack of commercially available thick walled high temperature alloys. Two alternative concepts, at reduced operating pressure, were investigated. Both were found to be readily fabricable to operate at 1800 F and capital cost estimates for these are included. A design concept, which is outside the scope of this study, was briefly considered

  7. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-09-01

    A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

  8. Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

    2016-11-01

    Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

  9. Design and preliminary results of a fuel flexible industrial gas turbine combustor

    Science.gov (United States)

    Novick, A. S.; Troth, D. L.; Yacobucci, H. G.

    1981-01-01

    The design characteristics are presented of a fuel tolerant variable geometry staged air combustor using regenerative/convective cooling. The rich/quench/lean variable geometry combustor is designed to achieve low NO(x) emission from fuels containing fuel bound nitrogen. The physical size of the combustor was calculated for a can-annular combustion system with associated operating conditions for the Allison 570-K engine. Preliminary test results indicate that the concept has the potential to meet emission requirements at maximum continuous power operation. However, airflow sealing and improved fuel/air mixing are necessary to meet Department of Energy program goals.

  10. Preliminary Design Concept for a Reactor-internal CRDM

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

    2013-01-01

    A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later

  11. Review of SFR Design Safety using Preliminary Regulatory PSA Model

    International Nuclear Information System (INIS)

    Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

    2013-01-01

    The major objective of this research is to develop a risk model for regulatory verification of the SFR design, and thereby, make sure that the SFR design is adequate from a risk perspective. In this paper, the development result of preliminary regulatory PSA model of SFR is discussed. In this paper, development and quantification result of preliminary regulatory PSA model of SFR is discussed. It was confirmed that the importance PDRC and ADRC dampers is significant as stated in the result of KAERI PSA model. However, the importance can be changed significantly depending on assumption of CCCG and CCF factor of PDRC and ADRC dampers. SFR (sodium-cooled fast reactor) which is Gen-IV nuclear energy system, is designed to accord with the concept of stability, sustainability and proliferation resistance. KALIMER-600, which is under development in Korea, includes passive safety systems (e. g. passive reactor shutdown, passive residual heat removal, and etc.) as well as active safety systems. Risk analysis from a regulatory perspective is needed to support the regulatory body in its safety and licensing review for SFR (KALIMER-600). Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and delay of the SFR licensing schedule that may be caused otherwise

  12. Design guide for calculating fluid damping for circular cylindrical structures

    International Nuclear Information System (INIS)

    Chen, S.S.

    1983-06-01

    Fluid damping plays an important role for structures submerged in fluid, subjected to flow, or conveying fluid. This design guide presents a summary of calculational procedures and design data for fluid damping for circular cylinders vibrating in quiescent fluid, crossflow, and parallel flow

  13. Preliminary Design Progress of the HCCR TBM for ITER testing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Park, Sung Dae; Kim, Dong Jun; Jin, Hyung Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) including the TBM-shield, which is called the TBM-set, to be tested in ITER, a Nuclear Facility INB-174. Through the conceptual design review (CDR), its design integrity was successfully demonstrated at the conceptual design level at various loads. After CD approval, preliminary design (PD) was started and the progress is introduced in the present study. After PD review and approval, final design and then fabrication will be started. The main purpose of PD is to design the TBM-set according to the fabrication aspect and more detailed design for interfaces with ITER machine, such as installed TBM port plug and frame. With these considering, PD of TBM-set was started. PD for HCCR TBM has been performed (so far v0.24) from the CD model. FW, BZ, SW, TES/NAS, BM, and connecting support design were performed through the analyses, if necessary. The manufacturability was the main concern for PD model development. Thermal hydraulic analysis will be performed to evaluate the temperature and pressure drop in TBM-set. The structural integrity of TBM-set will be confirmed with combined various loads condition.

  14. Preliminary regulatory audit calculation for Shinkori Units 3 and 4 LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Woo, S. W.; Kim, B. S.; Kim, J. K. (and others)

    2006-12-15

    The objective of this study is to perform a preliminary evaluation for Shinkori Units 3 and 4 LBLOCA by applying KINS Realistic Evaluation Methodology (REM). The following results were obtained: (1) From the evaluation for Shinkori Units 3 and 4 LBLOCA, the peak cladding temperature was evaluated to meet the regulatory requirement and the feasibility of the KINS-REM was identified. (2) The input decks that were developed in the previous studies, were reviewed and the evaluation model of the fluidic device was developed and applied for the audit calculation. (3) The treating method for the uncertainty of the gap conductance was developed and applied for the audit calculation. (4) The pre- and post-processing programs were developed for this study. (5) For the more detailed assessments, the information for the gap conductance, etc. should be improved and the effects of coolant bypass during blowdown, steam binding and so on were not sufficiently evaluated. KINS-REM should be advanced to evaluate these effects properly. The KINS methodology that was used in this study, can be further applied for independent regulatory audit calculations related to the licensing application on LOCA best estimate calculation.

  15. Preliminary Design of Reluctance Motors for Light Electric Vehicles Driving

    Directory of Open Access Journals (Sweden)

    TRIFA, V.

    2009-02-01

    Full Text Available The paper presents the aspects regarding FEM analysis of a reluctant motor for direct driving of the light electric vehicles. The reluctant motor take into study is of special construction suitable for direct drive of a light electric vehicle. It is an inverse radial reluctant motor, with a fixed stator mounted on front wheel shaft and an external toothed rotor fixed on the front wheel itself. A short presentation of preliminary design is continued with the FEM analysis in order to provide the optimal geometry of the motor and adequate windings.

  16. Preliminary shielding design evaluation for reactor assembly of SMART

    International Nuclear Information System (INIS)

    Kim, Kyo Youn; Kang, Chang M.; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    This report describes a preliminary evaluations of SMART shielding design near the reactor core by using the DORT two-dimensional discrete ordinates transport code. The results indicate that maximum neutron fluence at the bottom of reactor vessel is 1.64x10 17 n/cm 2 and that on the radial surface of reactor vessel is 6.71x10 16 n/cm 2 . These results meet the requirement, 1.0x10 20 n/cm 2 , in 10 CFR 50.61 and the integrity of SMART reactor vessel is confirmed during the lifetime of reactor. (Author). 20 refs., 11 tabs., 8 figs

  17. Gas cooled fast reactor 2400 MWTh, status on the conceptual design studies and preliminary safety analysis

    International Nuclear Information System (INIS)

    Malo, J.Y.; Alpy, N.; Bentivoglio, F.

    2009-01-01

    The Gas cooled Fast Reactor (GFR) is considered by the French Commissariat a l'Energie Atomique as a promising concept, combining the benefits of fast spectrum and high temperature, using Helium as coolant. A status on the GFR preliminary viability was made at the end of 2007, ending the pre-conceptual design phase. A consistent overall systems arrangement was proposed and a preliminary safety analysis based on operating transient calculations and a simplified PSA had established a global confidence in the feasibility and safety of this baseline concept. Its potential for attractive performances had been pointed out. Compare to the more mature Sodium Fast Reactor technology, no demonstrator has ever been built and the feasibility demonstration will required a longer lead time. The next main project milestone is related to the GFR viability, scheduled in 2012. The current studies consist in revisiting the reactor reference design options as selected at the end of 2007. Most of them are being consolidated by going more in depth in the analysis. Some possible alternatives are assessed. The paper will give a status on the last studies performed on the core design and corresponding neutronics and cycle performance, the Decay Heat Removal strategy and preliminary safety analysis, systems design and balance of plant... This paper is complementary to the Icapp'09 papers 9062 dealing with the Gas cooled Fast Reactor Demonstrator ALLEGRO and 9378 related to GFR transients analysis. (author)

  18. Preliminary design and off-design performance analysis of an Organic Rankine Cycle for geothermal sources

    International Nuclear Information System (INIS)

    Hu, Dongshuai; Li, Saili; Zheng, Ya; Wang, Jiangfeng; Dai, Yiping

    2015-01-01

    Highlights: • A method for preliminary design and performance prediction is established. • Preliminary data of radial inflow turbine and plate heat exchanger are obtained. • Off-design performance curves of critical components are researched. • Performance maps in sliding pressure operation are illustrated. - Abstract: Geothermal fluid of 90 °C and 10 kg/s can be exploited together with oil in Huabei Oilfield of China. Organic Rankine Cycle is regarded as a reasonable method to utilize these geothermal sources. This study conducts a detailed design and off-design performance analysis based on the preliminary design of turbines and heat exchangers. The radial inflow turbine and plate heat exchanger are selected in this paper. Sliding pressure operation is applied in the simulation and three parameters are considered: geothermal fluid mass flow rate, geothermal fluid temperature and condensing pressure. The results indicate that in all considered conditions the designed radial inflow turbine has smooth off-design performance and no choke or supersonic flow are found at the nozzle and rotor exit. The lager geothermal fluid mass flow rate, the higher geothermal fluid temperature and the lower condensing pressure contribute to the increase of cycle efficiency and net power. Performance maps are illustrated to make system meet different load requirements especially when the geothermal fluid temperature and condensing pressure deviate from the design condition. This model can be used to provide basic data for future detailed design, and predict off-design performance in the initial design phase

  19. Design of a transport calculation system for logging sondes simulation

    International Nuclear Information System (INIS)

    Marquez Damian, Jose Ignacio

    2005-01-01

    Analysis of available resources in earth crust is performed by different techniques, one of them is neutron logging. Design of sondes that are used to make such logging is supported by laboratory experiments as well as by numerical calculations.This work presents several calculation schemes, designed to simplify the task of whom has to planify such experiments or optimize parameters of this kind of sondes.These schemes use transport calculation codes, especially DaRT, TORT and MCNP, and cross section processing modules from SCALE system.Additionally a system for DaRT and TORT data postprocessing using OpenDX is presented.It allows scalar flux spatial distribution analysis, as wells as cross section condensation and reaction rates calculation

  20. Space Launch Systems Block 1B Preliminary Navigation System Design

    Science.gov (United States)

    Oliver, T. Emerson; Park, Thomas; Anzalone, Evan; Smith, Austin; Strickland, Dennis; Patrick, Sean

    2018-01-01

    NASA is currently building the Space Launch Systems (SLS) Block 1 launch vehicle for the Exploration Mission 1 (EM-1) test flight. In parallel, NASA is also designing the Block 1B launch vehicle. The Block 1B vehicle is an evolution of the Block 1 vehicle and extends the capability of the NASA launch vehicle. This evolution replaces the Interim Cryogenic Propulsive Stage (ICPS) with the Exploration Upper Stage (EUS). As the vehicle evolves to provide greater lift capability, increased robustness for manned missions, and the capability to execute more demanding missions so must the SLS Integrated Navigation System evolved to support those missions. This paper describes the preliminary navigation systems design for the SLS Block 1B vehicle. The evolution of the navigation hard-ware and algorithms from an inertial-only navigation system for Block 1 ascent flight to a tightly coupled GPS-aided inertial navigation system for Block 1B is described. The Block 1 GN&C system has been designed to meet a LEO insertion target with a specified accuracy. The Block 1B vehicle navigation system is de-signed to support the Block 1 LEO target accuracy as well as trans-lunar or trans-planetary injection accuracy. Additionally, the Block 1B vehicle is designed to support human exploration and thus is designed to minimize the probability of Loss of Crew (LOC) through high-quality inertial instruments and robust algorithm design, including Fault Detection, Isolation, and Recovery (FDIR) logic.

  1. Radiology workstation for mammography: preliminary observations, eyetracker studies, and design

    Science.gov (United States)

    Beard, David V.; Johnston, Richard E.; Pisano, Etta D.; Hemminger, Bradley M.; Pizer, Stephen M.

    1991-07-01

    For the last four years, the UNC FilmPlane project has focused on constructing a radiology workstation facilitating CT interpretations equivalent to those with film and viewbox. Interpretation of multiple CT studies was originally chosen because handling such large numbers of images was considered to be one of the most difficult tasks that could be performed with a workstation. The authors extend the FilmPlane design to address mammography. The high resolution and contrast demands coupled with the number of images often cross- compared make mammography a difficult challenge for the workstation designer. This paper presents the results of preliminary work with workstation interpretation of mammography. Background material is presented to justify why the authors believe electronic mammographic workstations could improve health care delivery. The results of several observation sessions and a preliminary eyetracker study of multiple-study mammography interpretations are described. Finally, tentative conclusions of what a mammographic workstation might look like and how it would meet clinical demand to be effective are presented.

  2. Ontario electricity industry restructuring : preliminary asset valuation and calculation of stranded debt

    International Nuclear Information System (INIS)

    1998-01-01

    The rationale for restructuring Ontario's electricity industry was restated. Financial elements of the Government's White Paper on the electrical industry included the following: (1) establishing a level playing field on taxes and regulation, (2) restructuring Ontario Hydro into new companies with clear business mandates, and (3) taking action to put the new companies on solid financial ground. To achieve these objectives requires valuation of the new companies as a key part in the restructuring process. This Ministry of Finance document contains preliminary estimates of the total debt and liabilities of Ontario Hydro ($ 39.1 billion), the value of the new generation and service companies ($ 15.8 billion), and the stranded debt ($ 23.3 billion, less the value of dedicated revenue streams of $ 15.4 billion, equal to the residual stranded debt of $ 7.9 billion). The method by which the stranded debt was calculated is also described. It is stressed that the overriding principles governing the financial restructuring plan are to achieve restructuring without increasing electricity rates, to retain maximum value in the electricity sector until stranded debt is retired, and to recover stranded debt from the electricity sector and not from taxpayers. Ministry advisors indicate that these preliminary valuations would allow the new companies to operate as commercial companies in a competitive market and receive investment grade credit ratings. 44 figs

  3. Preliminary study of magnet design for an SSC

    International Nuclear Information System (INIS)

    Taylor, C.E.; Meuser, R.B.

    1983-08-01

    The overriding design consideration for the SSC magnets is that cost of the facility be minimized; at 8 T, approximately 40 km of bending magnets is required for each ring of a 20 TeV collider. We present some results of a parametric study of two-in-one, iron-core magnets for an SSC. These results are necessarily preliminary in nature, and are intended only to show some of the trade-offs for a wide range of the variables. We show also some results for a reference design that produces 6.5 T in the aperture at 4.4 K for a coil inside diameter of 40 mm. It is not to be inferred that we have established this to be an optimum in any sense

  4. Preliminary design of a dedicated proton therapy linac

    International Nuclear Information System (INIS)

    Hamm, R.W.; Crandall, K.R.; Potter, J.M.

    1991-01-01

    The preliminary design has been completed for a low current, compact proton linac dedicated to cancer therapy. A 3 GHz side-coupled structure accelerates the beam from a 70 MeV drift tube linac using commercially available S-band rf power systems and accelerating cavities. This significantly reduces the linac cost and allows incremental energies up to 250 MeV. The short beam pulse width and high repetition rate make the linac similar to the high energy electron linacs now used for cancer therapy, yet produce a proton flux sufficient for treatment of large tumors. The high pulse repetition rate permits raster scanning, and the small output beam size and emittance result in a compact isocentric gantry design. Such a linac will reduce the facility and operating costs for a dedicated cancer therapy system

  5. Preliminary design of the advanced quantum beam source

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Cheol; Lee, Jong Min; Jeong, Young Uk; Cho, Sung Oh; Yoo, Jae Gwon; Park, Seong Hee

    2000-07-01

    The preliminary design of the advanced quantum beam source based on a superconducting electron accelerator is presented. The advanced quantum beams include: high power free electron lasers, monochromatic X-rays and {gamma}-rays, high-power medium-energy electrons, high-flux pulsed neutrons, and high-flux monochromatic slow positron beam. The AQBS system is being re-designed, assuming that the SPS superconducting RF cavities used for LEP at CERN will revived as a main accelerator of the AQBS system at KAERI, after the decommissioning of LEP at the end of 2000. Technical issues of using the SPS superconducting RF cavities for the AQBS project are discussed in this report. The advanced quantum beams will be used for advanced researches in science and industries.

  6. Preliminary design of the thermal protection system for solar probe

    Science.gov (United States)

    Dirling, R. B., Jr.; Loomis, W. C.; Heightland, C. N.

    1982-01-01

    A preliminary design of the thermal protection system for the NASA Solar Probe spacecraft is presented. As presently conceived, the spacecraft will be launched by the Space Shuttle on a Jovian swing-by trajectory and at perihelion approach to three solar radii of the surface of the Earth's sun. The system design satisfies maximum envelope, structural integrity, equipotential, and mass loss/contamination requirements by employing lightweight carbon-carbon emissive shields. The primary shield is a thin shell, 15.5-deg half-angle cone which absorbs direct solar flux at up to 10-deg off-nadir spacecraft pointing angles. Secondary shields of sandwich construction and low thickness-direction thermal conductivity are used to reduce the primary shield infrared radiation to the spacecraft payload.

  7. Preliminary design of the advanced quantum beam source

    International Nuclear Information System (INIS)

    Lee, Byung Cheol; Lee, Jong Min; Jeong, Young Uk; Cho, Sung Oh; Yoo, Jae Gwon; Park, Seong Hee

    2000-07-01

    The preliminary design of the advanced quantum beam source based on a superconducting electron accelerator is presented. The advanced quantum beams include: high power free electron lasers, monochromatic X-rays and γ-rays, high-power medium-energy electrons, high-flux pulsed neutrons, and high-flux monochromatic slow positron beam. The AQBS system is being re-designed, assuming that the SPS superconducting RF cavities used for LEP at CERN will revived as a main accelerator of the AQBS system at KAERI, after the decommissioning of LEP at the end of 2000. Technical issues of using the SPS superconducting RF cavities for the AQBS project are discussed in this report. The advanced quantum beams will be used for advanced researches in science and industries

  8. Preliminary Design Through Graphs: A Tool for Automatic Layout Distribution

    Directory of Open Access Journals (Sweden)

    Carlo Biagini

    2015-02-01

    Full Text Available Diagrams are essential in the preliminary stages of design for understanding distributive aspects and assisting the decision-making process. By drawing a schematic graph, designers can visualize in a synthetic way the relationships between many aspects: functions and spaces, distribution of layouts, space adjacency, influence of traffic flows within a facility layout, and so on. This process can be automated through the use of modern Information and Communication Technologies tools (ICT that allow the designers to manage a large quantity of information. The work that we will present is part of an on-going research project into how modern parametric software influences decision-making on the basis of automatic and optimized layout distribution. The method involves two phases: the first aims to define the ontological relation between spaces, with particular reference to a specific building typology (rules of aggregation of spaces; the second entails the implementation of these rules through the use of specialist software. The generation of ontological relations begins with the collection of data from historical manuals and analyses of case studies. These analyses aim to generate a “relationship matrix” based on preferences of space adjacency. The phase of implementing the previously defined rules is based on the use of Grasshopper to analyse and visualize different layout configurations. The layout is generated by simulating a process involving the collision of spheres, which represents specific functions of the design program. The spheres are attracted or rejected as a function of the relationships matrix, as defined above. The layout thus obtained will remain in a sort of abstract state independent of information about the exterior form, but will still provide a useful tool for the decision-making process. In addition, preliminary results gathered through the analysis of case studies will be presented. These results provide a good variety

  9. Optimized design for TWR assembly by CFD calculations

    International Nuclear Information System (INIS)

    Lu Jianchao; Lu Chuan; Yan Mingyu

    2013-01-01

    High temperature difference in travelling wave reactor bundle was found in the previous work. It could not be used in bundle design. Various analysis focused on helical wrapped wires and assembly housing was carried out by CFD calculation which found that the helical wrapped wires could influence the temperature differences while the effect was not obvious. Adding the strips and fillets on the assembly housing could optimize the thermal characteristics greatly, which can be used in the TWR assembly design. (authors)

  10. Approach to equilibrium calculations for the dragon HTR design

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1971-06-10

    The calculational methods and the model used in representing the core and the fuel management operations are described. Different layouts of the first core and approach to equilibrium schemes for the Dragon HTR design are investigated. A simple fuelling modus is found and the tchnological and economical implications are discussed in detail.

  11. Design and thermal-hydraulic calculation for EAST PFCs' baking

    International Nuclear Information System (INIS)

    Wan Xiaogang; Yao Damao

    2006-01-01

    According to the vacuum requirements for fusion in a tokamak device, the authors adopted a kind of gas flow baking technique in EAST. This paper presented the sketch design for EAST PFCs' baking, selected the specifications for the working gas. Calculated the hydraulic and thermal conditions in PFCs under baking, and simulated the results. (authors)

  12. Preliminary Design Optimization For A Supersonic Turbine For Rocket Propulsion

    Science.gov (United States)

    Papila, Nilay; Shyy, Wei; Griffin, Lisa; Huber, Frank; Tran, Ken; McConnaughey, Helen (Technical Monitor)

    2000-01-01

    In this study, we present a method for optimizing, at the preliminary design level, a supersonic turbine for rocket propulsion system application. Single-, two- and three-stage turbines are considered with the number of design variables increasing from 6 to 11 then to 15, in accordance with the number of stages. Due to its global nature and flexibility in handling different types of information, the response surface methodology (RSM) is applied in the present study. A major goal of the present Optimization effort is to balance the desire of maximizing aerodynamic performance and minimizing weight. To ascertain required predictive capability of the RSM, a two-level domain refinement approach has been adopted. The accuracy of the predicted optimal design points based on this strategy is shown to he satisfactory. Our investigation indicates that the efficiency rises quickly from single stage to 2 stages but that the increase is much less pronounced with 3 stages. A 1-stage turbine performs poorly under the engine balance boundary condition. A portion of fluid kinetic energy is lost at the turbine discharge of the 1-stage design due to high stage pressure ratio and high-energy content, mostly hydrogen, of the working fluid. Regarding the optimization technique, issues related to the design of experiments (DOE) has also been investigated. It is demonstrated that the criteria for selecting the data base exhibit significant impact on the efficiency and effectiveness of the construction of the response surface.

  13. Preliminary design of the beam transport system for the Milan biomedical cyclotron

    International Nuclear Information System (INIS)

    Silari, M.

    1988-01-01

    This report illustrates the preliminary design of the beam transport system for the Scanditronix MC40 cyclotron to be installed in Milan. The Cyclotron will be dedicated to biomedical research and the different experimental conditions that could occur will require a beam transport system flexible enough so as to deliver beams with the specified characteristics. The report describes the computer codes used, the calculations performed and the results obtained. The complete configuration of the beam lines serving the first two target rooms is given, together with typical beam profiles and the emittance ellipse variation along the transfer channels

  14. theory and calculation of the design of nuclear reactor

    International Nuclear Information System (INIS)

    Refaat, R.A.

    1994-01-01

    For the sake of formation of a complete general code for nuclear power reactor design, this thesis deals with a great part of this code. the code links the solution of the neutron integral transport equation by the multigroup treatment (76 energy groups) for the calculation of the reactor cell parameters by the fuel management program that solves the neutron diffusion equation inside a large number of nuclear fuel assemblies. the lattice cell code is modified to accommodate the calculation of lattice cell parameters for more than one enrichment ( one after the other). it is also modified to calculate the burn up parameters using unequal time steps. these two modifications are complicated but necessary for the link between the cell program and fuel management program. the comparison between the results of the fitted cross sections and that given by the cell calculations shows the necessity of using the cell code cross sections. this is also necessary for the sake of generality for any type of reactors. the comparison for the fuel management calculation depending on fitted data and that depending on cell calculation data insures the necessity for using the cell data i.e. insures the necessity of linking the cell calculation program by the fuel management program

  15. Design and structural calculation of nuclear power plant mechanical components

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do

    1986-01-01

    The mechanical components of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design. In this paper, it is intended to describe the design and structural calculations concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities. (Author) [pt

  16. Preliminary Design Study of the Hollow Electron Lens for LHC

    CERN Document Server

    Perini, Diego; CERN. Geneva. ATS Department

    2017-01-01

    A Hollow Electron Lens (HEL) has been proposed in order to improve performance of halo control and collimation in the Large Hadron Collider in view of its High Luminosity upgrade (HL-LHC). The concept is based on a beam of electrons that travels around the protons for a few meters. The electron beam is produced by a cathode and then guided by a strong magnetic field generated by a set of superconducting solenoids. The first step of the design is the definition of the magnetic fields that drive the electron trajectories. The estimation of such trajectories by means of a dedicated MATLAB® tool is presented. The influence of the main geometrical and electrical parameters are analysed and discussed. Then, the main mechanical design choices for the solenoids, cryostats gun and collector are described. The aim of this paper is to provide an overview of the preliminary design of the Electron Lens for LHC. The methods used in this study also serve as examples for future mechanical and integration designs of similar ...

  17. Ultraviolet Free Electron Laser Facility preliminary design report

    Energy Technology Data Exchange (ETDEWEB)

    Ben-Zvi, I. (ed.)

    1993-02-01

    This document, the Preliminary Design Report (PDR) for the Brookhaven Ultraviolet Free Electron Laser (UV FEL) facility, describes all the elements of a facility proposed to meet the needs of a research community which requires ultraviolet sources not currently available as laboratory based lasers. Further, for these experiments, the requisite properties are not extant in either the existing second or upcoming third generation synchrotron light sources. This document is the result of our effort at BNL to identify potential users, determine the requirements of their experiments, and to design a facility which can not only satisfy the existing need, but have adequate flexibility for possible future extensions as need dictates and as evolving technology allows. The PDR is comprised of three volumes. In this, the first volume, background for the development of the proposal is given, including descriptions of the UV FEL facility, and representative examples of the science it was designed to perform. Discussion of the limitations and potential directions for growth are also included. A detailed description of the facility design is then provided, which addresses the accelerator, optical, and experimental systems. Information regarding the conventional construction for the facility is contained in an addendum to volume one (IA).

  18. Ultraviolet Free Electron Laser Facility preliminary design report

    International Nuclear Information System (INIS)

    Ben-Zvi, I.

    1993-02-01

    This document, the Preliminary Design Report (PDR) for the Brookhaven Ultraviolet Free Electron Laser (UV FEL) facility, describes all the elements of a facility proposed to meet the needs of a research community which requires ultraviolet sources not currently available as laboratory based lasers. Further, for these experiments, the requisite properties are not extant in either the existing second or upcoming third generation synchrotron light sources. This document is the result of our effort at BNL to identify potential users, determine the requirements of their experiments, and to design a facility which can not only satisfy the existing need, but have adequate flexibility for possible future extensions as need dictates and as evolving technology allows. The PDR is comprised of three volumes. In this, the first volume, background for the development of the proposal is given, including descriptions of the UV FEL facility, and representative examples of the science it was designed to perform. Discussion of the limitations and potential directions for growth are also included. A detailed description of the facility design is then provided, which addresses the accelerator, optical, and experimental systems. Information regarding the conventional construction for the facility is contained in an addendum to volume one (IA)

  19. Shielding calculation techniques used in the design of storage systems

    International Nuclear Information System (INIS)

    Wang, S.S.; Massey, J.V.

    1986-01-01

    The shielding design and analysis of a concrete modular spent fuel storage system are discussed. Particular attention is given to comparing various computation techniques in determining bulk shielding thickness, and also in dealing with the radiation streaming effect through the air exist penetration openings in the module. Three computer codes QADMOD, ANISN, and DOT-IV were used to solve the same problem. In addition, hand albedo calculation were done to augment the result of the QADMOD calculation to properly deal with the surface scattering

  20. A preliminary design of the collinear dielectric wakefield accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Zholents, A.; Gai, W.; Doran, S.; Lindberg, R.; Power, J.G.; Strelnikov, N.; Sun, Y.; Trakhtenberg, E.; Vasserman, I. [ANL, Argonne, IL 60439 (United States); Jing, C.; Kanareykin, A.; Li, Y. [Euclid Techlabs LLC, Solon, OH 44139 (United States); Gao, Q. [Tsinghua University, Beijing (China); Shchegolkov, D.Y.; Simakov, E.I. [LANL, Los Alamos, NM 87545 (United States)

    2016-09-01

    A preliminary design of the multi-meter long collinear dielectric wakefield accelerator that achieves a highly efficient transfer of the drive bunch energy to the wakefields and to the witness bunch is considered. It is made from ~0.5 m long accelerator modules containing a vacuum chamber with dielectric-lined walls, a quadrupole wiggler, an rf coupler, and BPM assembly. The single bunch breakup instability is a major limiting factor for accelerator efficiency, and the BNS damping is applied to obtain the stable multi-meter long propagation of a drive bunch. Numerical simulations using a 6D particle tracking computer code are performed and tolerances to various errors are defined.

  1. Preliminary design for a pierce wiggler beamstick and addendum

    International Nuclear Information System (INIS)

    Pirkle, D.

    1988-05-01

    Lawrence Livermore National Laboratory is developing a fast tunable microwave source for operation at 250 GHz and 10kW peak output power. This report presents the preliminary design of a Pierce gun and solenoid magnet that will be compatible with a Pierce-wiggler electron beam formation system (beamstick). The beamstick will be an appropriate power source for a tunable gyro-BWO at 250 GHz. Figure 1 presents the major components of the Pierce-wiggler beamstick: the electron gun, solenoid, beam tunnel, wiggler, and vacuum valve. Figure 2 shows an artistic conception of how the beamstick will interface with the interaction magnet, modulator and gyro-BWO circuit at MIT. 15 figs

  2. The Pierre Auger Observatory Upgrade - Preliminary Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Aab, Alexander [Univ. Siegen (Germany); et al.

    2016-04-12

    The Pierre Auger Observatory has begun a major Upgrade of its already impressive capabilities, with an emphasis on improved mass composition determination using the surface detectors of the Observatory. Known as AugerPrime, the upgrade will include new 4 m2 plastic scintillator detectors on top of all 1660 water-Cherenkov detectors, updated and more flexible surface detector electronics, a large array of buried muon detectors, and an extended duty cycle for operations of the fluorescence detectors. This Preliminary Design Report was produced by the Collaboration in April 2015 as an internal document and information for funding agencies. It outlines the scientific and technical case for AugerPrime. We now release it to the public via the arXiv server. We invite you to review the large number of fundamental results already achieved by the Observatory and our plans for the future.

  3. Calculation of drift seepage for alternative emplacement designs

    International Nuclear Information System (INIS)

    Li, Guomin; Tsang, Chin-Fu; Birkholzer, Jens

    1999-01-01

    The calculations presented in this report are performed to obtain seepage rates into drift and boreholes for two alternative designs of drift and waste emplacement at Yucca Mountain. The two designs are defined according to the Scope of Work 14012021M1, activity 399621, drafted October 6, 1998, and further refined in a conference telephone call on October 13, 1998, between Mark Balady, Jim Blink, Rob Howard and Chin-Fu Tsang. The 2 designs considered are: (1) Design A--Horizontal boreholes 1.0 m in diameter on both sides of the drift, with each borehole 8 m long and inclined to the drift axis by 30 degrees. The pillar between boreholes, measured parallel to the drift axis, is 3.3 m. In the current calculations, a simplified model of an isolated horizontal borehole 8 m long will be simulated. The horizontal borehole will be located in a heterogeneous fracture continuum representing the repository layer. Three different realizations will be taken from the heterogeneous field, representing three different locations in the rock. Seepage for each realization is calculated as a function of the percolation flux. Design B--Vertical boreholes, 1.0 m in diameter and 8.0 m deep, drilled from the bottom of an excavated 8.0 m diameter drift. Again, the drift with the vertical borehole will be assumed to be located in a heterogeneous fracture continuum, representing the rock at the repository horizon. Two realizations are considered, and seepage is calculated for the 8-m drift with and without the vertical 1-m borehole at its bottom

  4. Greenridge Multi-Pollutant Control Project Preliminary Public Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Connell, Daniel P

    2009-01-12

    the commercial readiness of an emissions control system that is specifically designed to meet the environmental compliance requirements of these smaller coal-fired EGUs. The multi-pollutant control system is being installed and tested on the AES Greenidge Unit 4 (Boiler 6) by a team including CONSOL Energy Inc. as prime contractor, AES Greenidge LLC as host site owner, and Babcock Power Environmental Inc. as engineering, procurement, and construction contractor. All funding for the project is being provided by the U.S. Department of Energy, through its National Energy Technology Laboratory, and by AES Greenidge. AES Greenidge Unit 4 is a 107 MW{sub e} (net), 1950s vintage, tangentially-fired, reheat unit that is representative of many of the 440 smaller coal-fired units identified above. Following design and construction, the multi-pollutant control system will be demonstrated over an approximately 20-month period while the unit fires 2-4% sulfur eastern U.S. bituminous coal and co-fires up to 10% biomass. This Preliminary Public Design Report is the first in a series of two reports describing the design of the multi-pollutant control facility that is being demonstrated at AES Greenidge. Its purpose is to consolidate for public use all available nonproprietary design information on the Greenidge Multi-Pollutant Control Project. As such, the report includes a discussion of the process concept, design objectives, design considerations, and uncertainties associated with the multi-pollutant control system and also summarizes the design of major process components and balance of plant considerations for the AES Greenidge Unit 4 installation. The Final Public Design Report, the second report in the series, will update this Preliminary Public Design Report to reflect the final, as-built design of the facility and to incorporate data on capital costs and projected operating costs.

  5. Preliminary conceptual design of target system. Pt. 1. System configuration

    Energy Technology Data Exchange (ETDEWEB)

    Hino, Ryutaro; Haga, Katsuhiro; Kaminaga, Masanori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1997-07-01

    In the 21st century, neutron is expected to play a very important role in the fields of structural biology, nuclear physics, material science if a very high-intensity neutron source will be built because of its superior nature as an probe to investigate material structure and its function. The Japan Atomic Energy Research Institute has launched the Neutron Science Project for construction and utilization of a high-intensity spallation neutron source coupled with a proton accelerator. In the project, a neutron scattering facility is planned to be constructed in an early stage. Development of a 5MW spallation neutron source is one of the most difficult technical challenges in this project. A two-step development plan of the target was established to construct a 5MW-target station In the 1st step, a 1.5MW target will be constructed to develop 5MW target technology. The preliminary conceptual design was conducted to clarify the specifications of the target system of 1.5MW and 5MW including system layout, scale etc. This report describes (1) a design policy, (2) a layout of system consisting of the target, remote-handling devices, bio-shieldings etc., (3) specifications of components and facilities such as cooling systems for target and moderators, beam-port shutter and air conditioning system, (4) overhaul procedures by remote-handling devices, (5) safety assessment, and (6) necessary R and D items derived from the design activity. (author)

  6. Preliminary site design for the SP-100 ground engineering test

    International Nuclear Information System (INIS)

    Cox, C.M.; Miller, W.C.; Mahaffey, M.K.

    1986-04-01

    In November, 1985, Hanford was selected by the Department of Energy (DOE) as the preferred site for a full-scale test of the integrated nuclear subsystem for SP-100. The Hanford Engineering Development Laboratory, operated by Westinghouse Hanford Company, was assigned as the lead contractor for the Test Site. The nuclear subsystem, which includes the reactor and its primary heat transport system, will be provided by the System Developer, another contractor to be selected by DOE in late FY-1986. In addition to reactor operations, test site responsibilities include preparation of the facility plus design, procurement and installation of a vacuum chamber to house the reactor, a secondary heat transport system to dispose of the reactor heat, a facility control system, and postirradiation examination. At the conclusion of the test program, waste disposal and facility decommissioning are required. The test site must also prepare appropriate environmental and safety evaluations. This paper summarizes the preliminary design requirements, the status of design, and plans to achieve full power operation of the test reactor in September, 1990

  7. Characterization of BPM pickup designs for the HESR rate at FAIR using simulations and numerical calculations

    Energy Technology Data Exchange (ETDEWEB)

    Halama, Arthur; Kamerdzhiev, Vsevolod; Boehme, Christian; Srinivasan, Sudharsan [Forschungszentrum Juelich, IKP-4 (Germany)

    2016-07-01

    The institute of Nuclear Physics 4(IKP-4) of the Research Center Juelich (FZJ) is in charge of building and commissioning the High Energy Storage Ring (HESR) within the international Facility for Antiproton and Ion Research (FAIR) at Darmstadt. Simulations and numerical calculations were performed to characterize the initial beam position pickup design. Capacitive couplings of the electrodes and the behavior of the electrical equivalent circuit were investigated. This made room for changes to the design and performance increase. A prototype of the BPM pickup was constructed and tested on a dedicated test bench. Preliminary results will be presented. In order to gain higher signal levels and higher sensitivity, another suggested design was characterized as well and put into comparison.

  8. The application of advanced rotor (performance) methods for design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Bussel, G.J.W. van [Delft Univ. of Technology, Inst. for Wind Energy, Delft (Netherlands)

    1997-08-01

    The calculation of loads and performance of wind turbine rotors has been a topic for research over the last century. The principles for the calculation of loads on rotor blades with a given specific geometry, as well as the development of optimal shaped rotor blades have been published in the decades that significant aircraft development took place. Nowadays advanced computer codes are used for specific problems regarding modern aircraft, and application to wind turbine rotors has also been performed occasionally. The engineers designing rotor blades for wind turbines still use methods based upon global principles developed in the beginning of the century. The question what to expect in terms of the type of methods to be applied in a design environment for the near future is addressed here. (EG) 14 refs.

  9. Coefficients Calculation in Pascal Approximation for Passive Filter Design

    Directory of Open Access Journals (Sweden)

    George B. Kasapoglu

    2018-02-01

    Full Text Available The recently modified Pascal function is further exploited in this paper in the design of passive analog filters. The Pascal approximation has non-equiripple magnitude, in contrast of the most well-known approximations, such as the Chebyshev approximation. A novelty of this work is the introduction of a precise method that calculates the coefficients of the Pascal function. Two examples are presented for the passive design to illustrate the advantages and the disadvantages of the Pascal approximation. Moreover, the values of the passive elements can be taken from tables, which are created to define the normalized values of these elements for the Pascal approximation, as Zverev had done for the Chebyshev, Elliptic, and other approximations. Although Pascal approximation can be implemented to both passive and active filter designs, a passive filter design is addressed in this paper, and the benefits and shortcomings of Pascal approximation are presented and discussed.

  10. Benchmark calculation of nuclear design code for HCLWR

    International Nuclear Information System (INIS)

    Suzuki, Katsuo; Saji, Etsuro; Gakuhari, Kazuhiko; Akie, Hiroshi; Takano, Hideki; Ishiguro, Yukio.

    1986-01-01

    In the calculation of the lattice cell for High Conversion Light Water Reactors, big differences of nuclear design parameters appear between the results obtained by various methods and nuclear data libraries. The validity of the calculation can be verified by the critical experiment. The benchmark calculation is also efficient for the estimation of the validity in wide range of lattice parameters and burnup. As we do not have many measured data. The benchmark calculations were done by JAERI and MAPI, using SRAC and WIMS-E respectively. The problem covered the wide range of lattice parameters, i.e., from tight lattice to the current PWR lattice. The comparison was made on the effective multiplication factor, conversion ratio, and reaction rate of each nuclide, including burnup and void effects. The difference of the result is largest at the tightest lattice. But even at that lattice, the difference of the effective multiplication factor is only 1.4 %. The main cause of the difference is the neutron absorption rate U-238 in resonance energy region. The difference of other nuclear design parameters and their cause were also grasped. (author)

  11. Piping and pipeline calculations manual construction, design fabrication and examination

    CERN Document Server

    Ellenberger, Philip

    2010-01-01

    The lack of commentary, or historical perspective, regarding the codes and standards requirements for piping design and construction is an obstacle to the designer, manufacturer, fabricator, supplier, erector, examiner, inspector, and owner who want to provide a safe and economical piping system. An intensive manual, this book will utilize hundreds of calculation and examples based on of 40 years of personal experiences of the author as both an engineer and instructor. Each example demonstrates how the code and standard has been correctly and incorrectly applied. This book is a ?no non

  12. Nuclear data sets for reactor design calculations - approved 1975

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  13. American National Standard: nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    1983-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include criteria for acceptance of evaluated nuclear data sets, criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  14. American National Standard nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    A standard is presented which identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  15. Validity test of design calculations of a PGNAA setup

    International Nuclear Information System (INIS)

    Naqvi, A.A.; Garwan, M.A.

    2004-01-01

    A rectangular moderator has been designed for the prompt gamma ray neutron activation analysis (PGNAA) setup at King Fahd University of Petroleum and Minerals (KFUPM) to analyze Portland cement samples. The design of the moderator assembly was obtained using Monte Carlo calculations. The design calculations of the new rectangular moderator of the KFUPM PGNAA setup have been verified experimentally through prompt gamma ray yield measurement as a function of the front moderator thickness. In this study the yield of the 3.54 and 4.94 MeV prompt gamma rays from silicon in a soil sample was measured as a function of thickness of the front moderator of the rectangular moderator. The experimental results were compared with the results of the Monte Carlo simulations. A good agreement has been achieved between the experimental results and the results of the calculations. The experimental results have provided useful information about the PGNAA setup performance, neutron moderation, and gamma ray attenuation in the PGNAA sample

  16. Preliminary neutronic design of spock reactor: A nuclear system for space power generation

    International Nuclear Information System (INIS)

    Burgio, N.; Santagata, A.; Cumo, M.; Fasano, A.; Frullini, M.

    2007-01-01

    Aim of this paper is to preliminary investigates the neutronic features of an upgrade of the MAUS [1] nuclear reactor whose core will be able to supply a thermoelectric converter in order to generate 30 kW of electricity for space applications. The neutronic layout of SPOCK (Space Power Core Ka) is a compact, MOX fuelled, liquid metal cooled and totally reflected fast reactor with a control system based on neutron absorption. Spock, that during the heart and launch operation must be maintained in sub-critical state, has to start up in the outer space at 40 K temperatures with the coolant in a solid state and it will reach the operating steady condition at the maximum temperature of 1300 K with the coolant in the liquid state. The main design goal is to maintains, in the operating conditions of a typical space mission, the control of the appropriate criticality margin versus temperature and coolant physical state. For this purpose, a neutronic/thermal-hydraulic calculation chain able to assists the entire design process must be set up. As preliminary recognition, MCNPX 2.5.0 and FLUENT calculations were carried out. The emerging key features of SPOCK are: an equilateral triangular mesh of 91 cylindrical UO 2 fuel rods with a Molybdenum clad ensured by two grids of the same material, cooled by liquid Sodium and contained in an AISI 316 L vessel. The core is totally wrapped by a Beryllium reflector that hosts six absorber (B 4 C) rotating control rods. The reactor shape is cylindrical (radius = 30 cm and height = 60 cm) with a total mass of 275 kg. The excess reactivity was of 5000 PCM at 1300 K. A preliminary evaluation of the control rods worth and a power spatial distribution were also discussed. Through the definition of an ideal reference K e ff value at 300 K for the actual SPOCK configuration, a sensitivity analysis on various cross sections data and material physical properties was performed for the given mission temperature range, allowing consideration on

  17. Design of a lube oil reservoir by using flow calculations

    Energy Technology Data Exchange (ETDEWEB)

    Rinkinen, J; Alfthan, A. [Institute of Hydraulics and Automation IHA, Tampere University of Technology, Tampere (Finland)] Suominen, J. [Institute of Energy and Process Engineering, Tampere University of Technology, Tampere (Finland); Airaksinen, A; Antila, K [R and D Engineer Safematic Oy, Muurame (Finland)

    1998-12-31

    The volume of usual oil reservoir for lubrication oil systems is designed by the traditional rule of thumb so that the total oil volume is theoretically changed in every 30 minutes by rated pumping capacity. This is commonly used settling time for air, water and particles to separate by gravity from the oil returning of the bearings. This leads to rather big volumes of lube oil reservoirs, which are sometimes difficult to situate in different applications. In this presentation traditionally sized lube oil reservoir (8 m{sup 3}) is modelled in rectangular coordinates and laminar oil flow is calculated by using FLUENT software that is based on finite difference method. The results of calculation are velocity and temperature fields inside the reservoir. The velocity field is used to visualize different particle paths through the reservoir. Particles that are studied by the model are air bubbles and water droplets. The interest of the study has been to define the size of the air bubbles that are released and the size of the water droplets that are separated in the reservoir. The velocity field is also used to calculate the modelled circulating time of the oil volume which is then compared with the theoretical circulating time that is obtained from the rated pump flow. These results have been used for designing a new lube oil reservoir. This reservoir has also been modelled and optimized by the aid of flow calculations. The best shape of the designed reservoir is constructed in real size for empirical measurements. Some results of the oil flow measurements are shown. (orig.) 7 refs.

  18. Design of a lube oil reservoir by using flow calculations

    Energy Technology Data Exchange (ETDEWEB)

    Rinkinen, J.; Alfthan, A. [Institute of Hydraulics and Automation IHA, Tampere University of Technology, Tampere (Finland)] Suominen, J. [Institute of Energy and Process Engineering, Tampere University of Technology, Tampere (Finland); Airaksinen, A.; Antila, K. [R and D Engineer Safematic Oy, Muurame (Finland)

    1997-12-31

    The volume of usual oil reservoir for lubrication oil systems is designed by the traditional rule of thumb so that the total oil volume is theoretically changed in every 30 minutes by rated pumping capacity. This is commonly used settling time for air, water and particles to separate by gravity from the oil returning of the bearings. This leads to rather big volumes of lube oil reservoirs, which are sometimes difficult to situate in different applications. In this presentation traditionally sized lube oil reservoir (8 m{sup 3}) is modelled in rectangular coordinates and laminar oil flow is calculated by using FLUENT software that is based on finite difference method. The results of calculation are velocity and temperature fields inside the reservoir. The velocity field is used to visualize different particle paths through the reservoir. Particles that are studied by the model are air bubbles and water droplets. The interest of the study has been to define the size of the air bubbles that are released and the size of the water droplets that are separated in the reservoir. The velocity field is also used to calculate the modelled circulating time of the oil volume which is then compared with the theoretical circulating time that is obtained from the rated pump flow. These results have been used for designing a new lube oil reservoir. This reservoir has also been modelled and optimized by the aid of flow calculations. The best shape of the designed reservoir is constructed in real size for empirical measurements. Some results of the oil flow measurements are shown. (orig.) 7 refs.

  19. Preliminary design studies on the Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Terry, W.J.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

    1992-08-01

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

  20. Preliminary design of the new Proton Synchrotron Internal Dump core

    CERN Document Server

    AUTHOR|(CDS)2091975; Nuiry, François-Xavier

    The luminosity of the LHC particle accelerator at CERN is planned to be upgraded in the first half of 2020s, requiring also the upgrade of its injector accelerators, including the Proton Synchrotron (PS). The PS Internal Dumps are beam dumps located in the PS accelerator ring. They are safety devices designed to stop the circulating proton beam in order to protect the accelerator from damage due to an uncontrolled beam loss. The PS Internal Dumps need to be upgraded to be able to withstand the future higher intensity and energy proton beams. The dump core is a block of material interacting with the beam. It is located in ultra-high vacuum and moved into the beam path in 150 milliseconds by an electromagnet and spring-based actuation mechanism. The circulating proton beam is shaved by the core surface during thousands of beam revolutions. The preliminary new dump core design weighs 13 kilograms and consists of an isostatically pressed fine-grain graphite and a precipitation hardened copper alloy CuCrZr. The ...

  1. Preliminary drift design analyses for nuclear waste repository in tuff

    International Nuclear Information System (INIS)

    Hardy, M.P.; Brechtel, C.E.; Goodrich, R.R.; Bauer, S.J.

    1990-01-01

    The Yucca Mountain Project (YMP) is examining the feasibility of siting a repository for high-level nuclear waste at Yucca Mountain, on and adjacent to the Nevada Test Site (NTS). The proposed repository will be excavated in the Topopah Spring Member, which is a moderately fractured, unsaturated, welded tuff. Excavation stability will be required during construction, waste emplacement, retrieval (if required), and closure to ensure worker safety. The subsurface excavations will be subject to stress changes resulting from thermal expansion of the rock mass and seismic events associated with regional tectonic activity and underground nuclear explosions (UNEs). Analyses of drift stability are required to assess the acceptable waste emplacement density, to design the drift shapes and ground support systems, and to establish schedules and cost of construction. This paper outlines the proposed methodology to assess drift stability and then focuses on an example of its application to the YMP repository drifts based on preliminary site data. Because site characterization activities have not begun, the database currently lacks the extensive site-specific field and laboratory data needed to form conclusions as to the final ground support requirements. This drift design methodology will be applied and refined as more site-specific data are generated and as analytical techniques and methodologies are verified during the site characterization process

  2. Permian Basin, Texas: Volume 1, Text: Final preliminary design report

    International Nuclear Information System (INIS)

    1988-01-01

    This report is a description of the preliminary design for an Exploratory Shaft Facility (ESF) at the proposed 49 acre site located 21 miles north of Hereford, Texas in Deaf Smith County. Department of Energy must conduct in situ testing at depth to ascertain the engineering and environmental suitability of the site for further consideration for nuclear waste repository development. The ESF includes the construction of two 12-ft diameter engineered shafts for accessing the bedded salt horizon to conduct in situ tests to ascertain if the site should be considered a candidate site for the first High Level Nuclear Waste Repository. This report includes pertinent engineering drawings for two shafts and all support facilities necessary for shaft construction and testing program operation. Shafts will be constructed by conventional drill-and-blast methods employing ground freezing prior to shaft construction to stabilize the existing groundwater and soil conditions at the site. A watertight liner and seal system will be employed to prevent intermingling of aquifers and provide a stable shaft throughout its design life. 38 refs., 37 figs., 14 tabs

  3. Virtual materials design using databases of calculated materials properties

    International Nuclear Information System (INIS)

    Munter, T R; Landis, D D; Abild-Pedersen, F; Jones, G; Wang, S; Bligaard, T

    2009-01-01

    Materials design is most commonly carried out by experimental trial and error techniques. Current trends indicate that the increased complexity of newly developed materials, the exponential growth of the available computational power, and the constantly improving algorithms for solving the electronic structure problem, will continue to increase the relative importance of computational methods in the design of new materials. One possibility for utilizing electronic structure theory in the design of new materials is to create large databases of materials properties, and subsequently screen these for new potential candidates satisfying given design criteria. We utilize a database of more than 81 000 electronic structure calculations. This alloy database is combined with other published materials properties to form the foundation of a virtual materials design framework (VMDF). The VMDF offers a flexible collection of materials databases, filters, analysis tools and visualization methods, which are particularly useful in the design of new functional materials and surface structures. The applicability of the VMDF is illustrated by two examples. One is the determination of the Pareto-optimal set of binary alloy methanation catalysts with respect to catalytic activity and alloy stability; the other is the search for new alloy mercury absorbers.

  4. Verification of EPA's ''Preliminary Remediation Goals for radionuclides'' (PRG) electronic calculator

    Energy Technology Data Exchange (ETDEWEB)

    Jannik, Tim [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Stagich, Brooke [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-28

    The U.S. Environmental Protection Agency (EPA) requested an external, independent verification study of their updated “Preliminary Remediation Goals for Radionuclides” (PRG) electronic calculator. The calculator provides PRGs for radionuclides that are used as a screening tool at Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and Resource Conservation and Recovery Act (RCRA) sites. These risk-based PRGs establish concentration limits under specific exposure scenarios. The purpose of this verification study is to determine that the calculator has no inherit numerical problems with obtaining solutions as well as to ensure that the equations are programmed correctly. There are 167 equations used in the calculator. To verify the calculator, all equations for each of seven receptor types (resident, construction worker, outdoor and indoor worker, recreator, farmer, and composite worker) were hand calculated using the default parameters. The same four radionuclides (Am-241, Co-60, H-3, and Pu-238) were used for each calculation for consistency throughout.

  5. A flexible Monte Carlo tool for patient or phantom specific calculations: comparison with preliminary validation measurements

    Science.gov (United States)

    Davidson, S.; Cui, J.; Followill, D.; Ibbott, G.; Deasy, J.

    2008-02-01

    The Dose Planning Method (DPM) is one of several 'fast' Monte Carlo (MC) computer codes designed to produce an accurate dose calculation for advanced clinical applications. We have developed a flexible machine modeling process and validation tests for open-field and IMRT calculations. To complement the DPM code, a practical and versatile source model has been developed, whose parameters are derived from a standard set of planning system commissioning measurements. The primary photon spectrum and the spectrum resulting from the flattening filter are modeled by a Fatigue function, cut-off by a multiplying Fermi function, which effectively regularizes the difficult energy spectrum determination process. Commonly-used functions are applied to represent the off-axis softening, increasing primary fluence with increasing angle ('the horn effect'), and electron contamination. The patient dependent aspect of the MC dose calculation utilizes the multi-leaf collimator (MLC) leaf sequence file exported from the treatment planning system DICOM output, coupled with the source model, to derive the particle transport. This model has been commissioned for Varian 2100C 6 MV and 18 MV photon beams using percent depth dose, dose profiles, and output factors. A 3-D conformal plan and an IMRT plan delivered to an anthropomorphic thorax phantom were used to benchmark the model. The calculated results were compared to Pinnacle v7.6c results and measurements made using radiochromic film and thermoluminescent detectors (TLD).

  6. Design and shielding calculation for a PET/CT facility

    International Nuclear Information System (INIS)

    Martin Escuela, J. M.; Palau San Pedro, A.; Lopez Diaz, A.

    2013-01-01

    Following the AAPM Task Group Report No. 108, the NCRP Report No. 147 recommendations and the Cuban's local regulations for nuclear medicine practice were carried out the safety planning and design of a new PET/CT facility for the Nuclear Medicine Department of 'Hermanos Ameijeiras' Hospital. It should be installed in the top floor of the NM building (3th floor), occupied by offices, classrooms and ancillaries areas, meanwhile in the second floor is working the conventional nuclear medicine department. The radiation doses were evaluated in areas of the second, third and quarter floor taking into account the pet isotope, the workload, the occupancy factors of each place, the use factors of different sources and the dose reduction factors, warranty the accomplish of the Cuban dose restrictions associated to the nuclear medicine practice. In each point of calculation was considered the contribution from each source to the total dose, as well as the contribution of the CT in the adjacent room to the imaging room. For the proper facility design was considered the transmission factors of the existing barriers, and calculated the new ones to be added between each source and the estimation point, keeping in mind the space limitations. The PET/CT design plan meet all the needs, the development of the project is consistent with the mission of the facility and the radiation protection regulations of nuclear medicine. (Author)

  7. THERMAL: A routine designed to calculate neutron thermal scattering

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1995-01-01

    THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy

  8. Understanding Creative Design Processes by Integrating Sketching and CAD Modelling Design Environments: A Preliminary Protocol Result from Architectural Designers

    Directory of Open Access Journals (Sweden)

    Yi Teng Shih

    2015-11-01

    Full Text Available This paper presents the results of a preliminary protocol study of the cognitive behaviour of architectural designers during the design process. The aim is to better understand the similarities and differences in cognitive behaviour using Sequential Mixed Media (SMM and Alternative Mixed Media (AMM approaches, and how switching between media may impact on design processes. Two participants with at least one-year’s professional design experience and a Bachelor of Design degree, and competence in both sketching and computer-aid design (CAD modelling participated in the study. Video recordings of participants working on different projects were coded using the Function-Behaviour-Structure (FBS coding scheme. Participants were also interviewed and their explanations about their switching behaviours were categorised into three types: S→C, S/C↹R and C→S. Preliminary results indicate that switching between media may influence how designers identify problems and develop solutions. In particular, two design issues were identified.  These relate to the FBS coding scheme, where structure (S and behaviour derived from structure (Bs, change to documentation (D after switching from sketching to CAD modelling (S→C. These switches make it possible for designers to integrate both approaches into one design medium and facilitate their design processes in AMM design environments.

  9. Subseabed radionuclide migration studies and preliminary repository design concepts

    International Nuclear Information System (INIS)

    Brush, L.H.

    1982-01-01

    Geochemical research carried out by the US Subseabed Disposal Program is described. Data from studies of high-temperature interactions between sediments and pore water (seawater) and from studies of sorption and diffusion of radionuclides in oxidized, deep-sea sediments are used, along with results from heat transfer studies, to predict migration rates of raionuclides in a subseabed repository. Preliminary results for most radionuclides in oxidized sediments are very encouraging. Fission products with moderate K/sub D/ values (10 2 to 10 5 ml/g) and actinides with high K/sub D/ values (10 3 to 10 6 ml/g) would not migrate significant distances before decaying to innocuous concentrations. Among this group are 137 Cs, 90 Sr, and 239 Pu. The results for anionic species in oxidized sediments are less encouraging. Planning for field verification of these laboratory and modeling studies is currently under way. Conceptual repository designs and emplacement options are also described. 33 references, 15 figures, 1 table

  10. Preliminary Design Report Shippingport Spent Fuel Drying and Inerting System

    International Nuclear Information System (INIS)

    JEPPSON, D.W.

    2000-01-01

    A process description and system flow sheets have been prepared to support the design/build package for the Shippingport Spent Fuel Canister drying and inerting process skid. A process flow diagram was prepared to show the general steps to dry and inert the Shippingport fuel loaded into SSFCs for transport and dry storage. Flow sheets have been prepared to show the flows and conditions for the various steps of the drying and inerting process. Calculations and data supporting the development of the flow sheets are included

  11. Radiological shielding of low power compact reactor: calculation and design

    International Nuclear Information System (INIS)

    Marino, Raul

    2004-01-01

    The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot [es

  12. A preliminary design of mechanical device on industrial digital radiography equipment design

    International Nuclear Information System (INIS)

    Nur Khasan; Samuel Praptoyo

    2015-01-01

    A preliminary design of mechanical device on industrial digital radiography equipment has been done. this design is intended as a basis for the manufacture of complete facilities for the realization a prototype on industrial digital radiography equipment. the design and construction were carried out by paying attention to the general configuration of the basic design in which its mechanical design has several components with specific dimensions and heavy mass. this design consist of a main frame holder, flat panel detector support and hydraulic hand stacker for mounting the x-ray machine. this mechanical device design will then be fabricated to facilitate and assist work of digital radiographic retrieval. computer application programs sketch-up is used to draw this design and the analysis stress of autodesk inventor to analysis the strength construction design. the results of this design are the configuration drawing, sketch drawings of construction and the safety factor of construction design with a minimum value of 2.39 as well as a maximum value of 15 when to be simulated by the load 500 Kg which is 4 times of total workload. (author)

  13. Extensions to the SCDAP/RELAP5 code for the modeling of debris oxidation and materials interactions preliminary design report

    International Nuclear Information System (INIS)

    Siefken, L.J.; Davis, K.L.

    1993-02-01

    Preliminary designs are proposed for extending the SCDAP/RELAP5 code so that it models (a) the oxidation of slumping fuel rod material and cohesive and porous debris and (b) the interaction of PWR control rod materials with the other materials in a reactor core. These extensions have the purpose of improving the code's calculation of the damage progression and hydrogen production that takes place during the early phase of a severe accident

  14. Risk-informed analysis as a support to the preliminary design of the CEA GFR2400

    International Nuclear Information System (INIS)

    Bertrand, F.; Bassi, C.; Azria, P.; Bentivoglio, F.; Messie, A.; Balmain, M.

    2012-01-01

    The integration of safety issues in the early phase of the design of a 4. generation reactor of the concepts is expected. For this purpose, probabilistic insights are increasingly employed in the safety demonstration in combination with the deterministic approach in the frame of a so-called risk informed approach. The present paper deals with the safety assessment of the preliminary design of the GFR2400 developed by CEA and how it has been improved in order to fulfil deterministic criteria as well as to reach a risk level comparable to the generation III reactors. GFR2400 is a 2400 MWth, 3-loops, helium-cooled fast reactor developed at a pre-conceptual design stage whose secondary circuit is filled with a mixture of helium and nitrogen, the ternary circuit being filled with water vaporized in 3 steam generators according to a classical Rankine cycle. The resulting cycle efficiency is very close to 45 %. Considering the results obtained with a preliminary level 1 PSA (L1PSA) model, it emerged that an increased reliability of the DHR (Decay Heat Removal) function in high pressure conditions (not corresponding to a LOCA) was suitable to reduce the overall core damage frequency. On the other hand, some small break LOCA situations were not adequately mitigated according to the line of protection deterministic method. Both issues have been solved by design improvements. In addition, this final L1PSA model, characterized by success criteria based on transient calculations performed with the CATHARE2 code and performed in a perimeter extended to all representative internal initiating events at full operating power, permitted to propose design evolutions that did not increase significantly the CDF. In the same time, those evolutions enabled the DHR system to increase its redundancy level as required in the deterministic approach. Finally, a modified design has been reached implying a more extended covering of various accidental situations by means of a progressive DHR

  15. Preliminary shielding calculation for the system of CyberKnife robotic radiosurgery

    International Nuclear Information System (INIS)

    Toreti, Dalila; Xavier, Clarice; Moura, Fabio

    2011-01-01

    The CyberKnife robotic system uses a manipulator with six grade of freedom for positioning a 6 MV Linac accelerator for treatment of lesions. This paper presents calculations for a standard room, with 200 cm of thickness walls primary, build for a CyberKnife system, and calculations for a room originally designed for a Linac conventional (with gantry), with secondary barriers of 107 cm thickness. After the realization of shielding for both rooms, the results shown that walls of standard room with 200 cm thickness are adequate for the secondary shield, and for a room with a conventional Linac, from all six evaluated points, two would require additional shielding of nine cm and four cm of concrete with 2.4 g/cubic cm. This shows that the CyberKnife system can be installed in a originally designed room for a conventional Linac with neither restrict nor any shielding, since no incidence of beams on the secondary barriers is existent

  16. Optimizing Parameters of Axial Pressure-Compounded Ultra-Low Power Impulse Turbines at Preliminary Design

    Science.gov (United States)

    Kalabukhov, D. S.; Radko, V. M.; Grigoriev, V. A.

    2018-01-01

    Ultra-low power turbine drives are used as energy sources in auxiliary power systems, energy units, terrestrial, marine, air and space transport within the confines of shaft power N td = 0.01…10 kW. In this paper we propose a new approach to the development of surrogate models for evaluating the integrated efficiency of multistage ultra-low power impulse turbine with pressure stages. This method is based on the use of existing mathematical models of ultra-low power turbine stage efficiency and mass. It has been used in a method for selecting the rational parameters of two-stage axial ultra-low power turbine. The article describes the basic features of an algorithm for two-stage turbine parameters optimization and for efficiency criteria evaluating. Pledged mathematical models are intended for use at the preliminary design of turbine drive. The optimization method was tested at preliminary design of an air starter turbine. Validation was carried out by comparing the results of optimization calculations and numerical gas-dynamic simulation in the Ansys CFX package. The results indicate a sufficient accuracy of used surrogate models for axial two-stage turbine parameters selection

  17. Inverse design-momentum, a method for the preliminary design of horizontal axis wind turbines

    International Nuclear Information System (INIS)

    Battisti, L; Soraperra, G; Fedrizzi, R; Zanne, L

    2007-01-01

    Wind turbine rotor prediction methods based on generalized momentum theory BEM routinely used in industry and vortex wake methods demand the use of airfoil tabulated data and geometrical specifications such as the blade spanwise chord distribution. They belong to the category of 'direct design' methods. When, on the other hand, the geometry is deduced from some design objective, we refer to 'inverse design' methods. This paper presents a method for the preliminary design of wind turbine rotors based on an inverse design approach. For this purpose, a generalized theory was developed without using classical tools such as BEM. Instead, it uses a simplified meridional flow analysis of axial turbomachines and is based on the assumption that knowing the vortex distribution and appropriate boundary conditions is tantamount to knowing the velocity distribution. The simple conservation properties of the vortex components consistently cope with the forces and specific work exchange expressions through the rotor. The method allows for rotor arbitrarily radial load distribution and includes the wake rotation and expansion. Radial pressure gradient is considered in the wake. The capability of the model is demonstrated first by a comparison with the classical actuator disk theory in investigating the consistency of the flow field, then the model is used to predict the blade planform of a commercial wind turbine. Based on these validations, the authors postulate the use of a different vortex distribution (i.e. not-uniform loading) for blade design and discuss the effect of such choices on blade chord and twist, force distribution and power coefficient. In addition to the method's straightforward application to the pre-design phase, the model clearly shows the link between blade geometry and performance allowing quick preliminary evaluation of non uniform loading on blade structural characteristics

  18. Georgetown University Integrated Community Energy System (GU-ICES). Phase III, Stage II. Preliminary design. Final report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-11-01

    Results are presented for two elements in the Georgetown University ICES program - the installation of a 2500-kW backpressure steam-turbine generator within a new extension to the heating and cooling plant (cogeneration) and the provision of four additional ash silos for the university's atmospheric fluidized-bed boiler plant (added storage scheme). The preliminary design and supporting documentation for the work items and architectural drawings are presented. Section 1 discusses the basis for the report, followed by sections on: feasibility analysis update; preliminary design documents; instrumentation and testing; revised work management plan; and appendices including outline constructions, turbine-generator prepurchase specification, design calculations, cost estimates, and Potomac Electric Company data. (MCW)

  19. Simplified methods and application to preliminary design of piping for elevated temperature service

    International Nuclear Information System (INIS)

    Severud, L.K.

    1975-01-01

    A number of simplified stress analysis methods and procedures that have been used on the FFTF project for preliminary design of piping operating at elevated temperatures are described. The rationale and considerations involved in developing the procedures and preliminary design guidelines are given. Applications of the simplified methods to a few FFTF pipelines are described and the success of these guidelines are measured by means of comparisons to pipeline designs that have had detailed Code type stress analyses. (U.S.)

  20. Current Status of HCCR TBM Design for the Preliminary Design Phase Preparation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Lee, Dong Won; Kim, Dong Jun [KAERI, Daejeon (Korea, Republic of); Ahn, Mu Young [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    Helium cooled ceramic reflector (HCCR) TBM-set will be installed in the equatorial port no.18 of ITER inside the vacuum vessel directly facing the plasma. TBM-set refers the TBM and associated shield and connecting support. After the Conceptual Design Review (CDR), Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) design is being updated for the preparation of the preliminary design phase. The manufacturability is considered based on the TBM-set model of the conceptual design phase. In this work, the design changes for each component of the TBM-set is described in comparison with the CD phase. The current design direction and details is presented. The first wall (FW) is component facing the plasma directly. This component should have a superior cooling performance. Present Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) design was described in comparison with the CD model. The manufacturability was considered in current PD phase. The detained design of the connecting support will be determined reflecting the load assessment. The structural integrity will be confirmed with a various load condition.

  1. Preliminary studi on neutronic aspect of a conceptual design of the Kartini reactor base ADS facility

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2012-01-01

    A preliminary study on neutronic aspect of a conceptual design of ADS facility with the basis of Kartini Reaktor, has been performed. The study was intended to see the feasibility from neutronic point of view of Kartini reactor, to be used as a small scale of NPP’s waste transmutation experimental facility. A SRAC code was used as the basis of calculations. The results indicate that the presence of minor actinides (MA) will give a positive reactivity, which tends to increase with the increase of MA concentrations. Based on the defined criteria of subcriticality and by considering the core power distributions and the level of reactivity contribution of MA element, it is concluded that Kartini reactor is potential enough to be used as an ADS experimental facility, mainly for MA concentration between 30 to 50 % of the assumed mixture of C-MA matrix. (author)

  2. Calculation of track and vertex errors for detector design studies

    International Nuclear Information System (INIS)

    Harr, R.

    1995-01-01

    The Kalman Filter technique has come into wide use for charged track reconstruction in high-energy physics experiments. It is also well suited for detector design studies, allowing for the efficient estimation of optimal track covariance matrices without the need of a hit level Monte Carlo simulation. Although much has been published about the Kalman filter equations, there is a lack of previous literature explaining how to implement the equations. In this paper, the operators necessary to implement the Kalman filter equations for two common detector configurations are worked out: a central detector in a uniform solenoidal magnetic field, and a fixed-target detector with no magnetic field in the region of the interactions. With the track covariance matrices in hand, vertex and invariant mass errors are readily calculable. These quantities are particularly interesting for evaluating experiments designed to study weakly decaying particles which give rise to displaced vertices. The optimal vertex errors are obtained via a constrained vertex fit. Solutions are presented to the constrained vertex problem with and without kinematic constraints. Invariant mass errors are obtained via propagation of errors; the use of vertex constrained track parameters is discussed. Many of the derivations are new or previously unpublished

  3. Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Zhou, E-mail: zhaozhou@swip.ac.cn; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

    2016-02-15

    Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li{sub 4}SiO{sub 4} pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

  4. Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

    International Nuclear Information System (INIS)

    Zhao, Zhou; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

    2016-01-01

    Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li_4SiO_4 pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

  5. A Preliminary Rubric Design to Evaluate Mixed Methods Research

    Science.gov (United States)

    Burrows, Timothy J.

    2013-01-01

    With the increase in frequency of the use of mixed methods, both in research publications and in externally funded grants there are increasing calls for a set of standards to assess the quality of mixed methods research. The purpose of this mixed methods study was to conduct a multi-phase analysis to create a preliminary rubric to evaluate mixed…

  6. OPAL shield design performance assessment. Comparison of measured dose rates against the corresponding design calculated values. A designer perspective

    Energy Technology Data Exchange (ETDEWEB)

    Brizuela, Martin; Albornoz, Felipe [INVAP SE, Av. Cmte. Piedrabuena, Bariloche (Argentina)

    2012-03-15

    A comparison of OPAL shielding calculations against measurements carried out during Commissioning, is presented for relevant structures such as the reactor block, primary shutters, neutron guide bunker, etc. All the results obtained agree very well with the measured values and contribute to establish the confidence on the calculation tools (MCNP4, DORT, etc.) and methodology used for shielding design. (author)

  7. Hierarchical modeling and robust synthesis for the preliminary design of large scale complex systems

    Science.gov (United States)

    Koch, Patrick Nathan

    Large-scale complex systems are characterized by multiple interacting subsystems and the analysis of multiple disciplines. The design and development of such systems inevitably requires the resolution of multiple conflicting objectives. The size of complex systems, however, prohibits the development of comprehensive system models, and thus these systems must be partitioned into their constituent parts. Because simultaneous solution of individual subsystem models is often not manageable iteration is inevitable and often excessive. In this dissertation these issues are addressed through the development of a method for hierarchical robust preliminary design exploration to facilitate concurrent system and subsystem design exploration, for the concurrent generation of robust system and subsystem specifications for the preliminary design of multi-level, multi-objective, large-scale complex systems. This method is developed through the integration and expansion of current design techniques: (1) Hierarchical partitioning and modeling techniques for partitioning large-scale complex systems into more tractable parts, and allowing integration of subproblems for system synthesis, (2) Statistical experimentation and approximation techniques for increasing both the efficiency and the comprehensiveness of preliminary design exploration, and (3) Noise modeling techniques for implementing robust preliminary design when approximate models are employed. The method developed and associated approaches are illustrated through their application to the preliminary design of a commercial turbofan turbine propulsion system; the turbofan system-level problem is partitioned into engine cycle and configuration design and a compressor module is integrated for more detailed subsystem-level design exploration, improving system evaluation.

  8. Liquid-liquid extraction. Choice, calculation and design of devices

    International Nuclear Information System (INIS)

    Leybros, J.

    2005-01-01

    The aim of this work is to study the problematic due to the choice of an equipment, to its size and to its industrial bringing into operation. Besides its efficiency to carry out a mass transfer, the economical interest of an industrial device will be proportional to its specific rate (volume treated by surface unit of the cross section of the extractor). Nevertheless, as it seems to be logic to maximize the three parameters which have an influence on it (transfer coefficient, interfacial surface and transfer potential), there exists no device which can modify separately any of these parameters. In order to satisfy these aims, a great diversity of devices have been put on the market. Indeed, it exists about twenty different industrial devices. This diversity leads the engineer, during the design of a solvent extraction device, to take difficult and subjective decisions without pilot experiment on the considered system. The main problem of an economical and optimized calculation of the liquid-liquid extractors from theoretical data has still not found totally satisfying solutions. Thus, except in some cases where we have empirical correlations, the recourse to prototype experiments is required and the most recent advances have resulted essentially to define more reliable rules for the extrapolation of these experiments for the size of industrial devices. (O.M.)

  9. Preliminary design and definition of field experiments for welded tuff rock mechanics program

    International Nuclear Information System (INIS)

    Zimmerman, R.M.

    1982-06-01

    The preliminary design contains objectives, typical experiment layouts, definitions of equipment and instrumentation, test matrices, preliminary design predictive modeling results for five experiments, and a definition of the G-Tunnel Underground Facility (GTUF) at the Nevada Test Site where the experiments are to be located. Experiments described for investigations in welded tuff are the Small Diameter Heater, Unit Cell-Canister Scale, Heated Block, Rocha Slot, and Miniature Heater

  10. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T. and others

    2006-01-01

    A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor

  11. Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

    1982-01-01

    Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

  12. Trade-off results and preliminary designs of Near-Term Hybrid Vehicles

    Science.gov (United States)

    Sandberg, J. J.

    1980-01-01

    Phase I of the Near-Term Hybrid Vehicle Program involved the development of preliminary designs of electric/heat engine hybrid passenger vehicles. The preliminary designs were developed on the basis of mission analysis, performance specification, and design trade-off studies conducted independently by four contractors. THe resulting designs involve parallel hybrid (heat engine/electric) propulsion systems with significant variation in component selection, power train layout, and control strategy. Each of the four designs is projected by its developer as having the potential to substitute electrical energy for 40% to 70% of the petroleum fuel consumed annually by its conventional counterpart.

  13. Design of software for calculation of shielding based on various standards radiodiagnostic calculation

    International Nuclear Information System (INIS)

    Falero, B.; Bueno, P.; Chaves, M. A.; Ordiales, J. M.; Villafana, O.; Gonzalez, M. J.

    2013-01-01

    The aim of this study was to develop a software application that performs calculation shields in radiology room depending on the type of equipment. The calculation will be done by selecting the user, the method proposed in the Guide 5.11, the Report 144 and 147 and also for the methodology given by the Portuguese Health Ministry. (Author)

  14. Efficient preliminary floating offshore wind turbine design and testing methodologies and application to a concrete spar design

    OpenAIRE

    Matha, Denis; Sandner, Frank; Molins i Borrell, Climent; Campos Hortigüela, Alexis; Cheng, Po Wen

    2015-01-01

    The current key challenge in the floating offshore wind turbine industry and research is on designing economic floating systems that can compete with fixed-bottom offshore turbines in terms of levelized cost of energy. The preliminary platform design, as well as early experimental design assessments, are critical elements in the overall design process. In this contribution, a brief review of current floating offshore wind turbine platform pre-design and scaled testing methodologies is provide...

  15. Preliminary seismic design cost-benefit assessment of the tuff repository waste-handling facilities

    International Nuclear Information System (INIS)

    Subramanian, C.V.; Abrahamson, N.; Hadjian, A.H.

    1989-02-01

    This report presents a preliminary assessment of the costs and benefits associated with changes in the seismic design basis of waste-handling facilities. The objectives of the study are to understand the capability of the current seismic design of the waste-handling facilities to mitigate seismic hazards, evaluate how different design levels and design measures might be used toward mitigating seismic hazards, assess the costs and benefits of alternative seismic design levels, and develop recommendations for possible modifications to the seismic design basis. This preliminary assessment is based primarily on expert judgment solicited in an interdisciplinary workshop environment. The estimated costs for individual attributes and the assumptions underlying these cost estimates (seismic hazard levels, fragilities, radioactive-release scenarios, etc.) are subject to large uncertainties, which are generally identified but not treated explicitly in this preliminary analysis. The major conclusions of the report do not appear to be very sensitive to these uncertainties. 41 refs., 51 figs., 35 tabs

  16. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

  17. Preliminary design of GDT-based 14 MeV neutron source

    International Nuclear Information System (INIS)

    Du Hongfei; Chen Dehong; Wang Hui; Wang Fuqiong; Jiang Jieqiong; Wu Yican; Chen Yiping

    2012-01-01

    To meet the need of D-T fusion neutron source for fusion material testing, design goals were presented in this paper according to the international requirements of neutron source for fusion material testing. A preliminary design scheme of GDT-based 14 MeV neutron source was proposed, and a physics model of the neutron source was built based on progress of GDT experiments. Two preliminary design schemes (i. e. FDS-GDT1, FDS-GDT2) were designed; among which FDS-GDT2 can be used for fusion material testing with neutron first wall loading of 2 MW/m 2 . (authors)

  18. Near-term hybrid vehicle program, phase 1. Appendix C: Preliminary design data package

    Science.gov (United States)

    1979-01-01

    The design methodology, the design decision rationale, the vehicle preliminary design summary, and the advanced technology developments are presented. The detailed vehicle design, the vehicle ride and handling and front structural crashworthiness analysis, the microcomputer control of the propulsion system, the design study of the battery switching circuit, the field chopper, and the battery charger, and the recent program refinements and computer results are presented.

  19. Preliminary study on flexible core design of super FBR with multi-axial fuel shuffling

    International Nuclear Information System (INIS)

    Sukarman; Yamaji, Akifumi; Someya, Takayuki; Noda, Shogo

    2017-01-01

    Preliminary study has been conducted on developing a new flexible core design concept for the Supercritical water-cooled Fast Breeder Reactor (Super FBR) with multi-axial fuel shuffling. The proposed new concept focuses on the characteristic large axial coolant density change in supercritical water cooled reactors (SCWRs) when the coolant inlet temperature is below the pseudocritical point and large coolant enthalpy rise is taken in the core for achieving high thermal efficiency. The aim of the concept is to attain both the high breeding performance and good thermal-hydraulic performance at the same time. That is, short Compound System Doubling Time (CSDT) for high breeding, large coolant enthalpy rise for high thermal efficiency, and large core power. The proposed core concept consists of horizontal layers of mixed oxide (MOX) fuels and depleted uranium (DU) blanket layers at different elevation levels. Furthermore, the upper core and the lower core are separated and independent fuel shuffling schemes in these two core regions are considered. The number of fuel batches and fuel shuffling scheme of the upper core were changed to investigate influence of multi-axial fuel shuffling on the core characteristics. The core characteristics are evaluated with-three-dimensional diffusion calculations, which are fully-coupled with thermal-hydraulics calculations based on single channel analysis model. The results indicate that the proposed multi-axial fuel shuffling scheme does have a large influence on CSDT. Further investigations are necessary to develop the core concept. (author)

  20. Failure mode analysis of preliminary design of ITER divertor impurity monitor

    International Nuclear Information System (INIS)

    Kitazawa, Sin-iti; Ogawa, Hiroaki

    2016-01-01

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • Failure mode of DIM was prepared for RAMI analysis. • RAMI analysis on DIM was performed to reduce technical risks. - Abstract: The objective of the divertor impurity influx monitor (DIM) for ITER is to measure the parameters of impurities and hydrogen isotopes (tritium, deuterium, and hydrogen) in divertor plasma using visible and UV spectroscopic techniques in the 200–1000 nm wavelength range. In ITER, special provisions are required to ensure accuracy and full functionality of the diagnostic components under harsh conditions (high temperature, high magnetic field, high vacuum condition, and high radiation field). Japan Domestic Agency is preparing the preliminary design of the ITER DIM system, which will be installed in the upper, equatorial and lower ports. The optical and mechanical designs of the DIM are conducted to fit ITER’s requirements. The optical and mechanical designs meet the requirements of spatial resolution. Some auxiliary systems were examined via prototyping. The preliminary design of the ITER DIM system was evaluated by RAMI analysis. The availability of the designed system is adequately high to satisfy the project requirements. However, some equipment does not have certain designs, and this may cause potential technical risks. The preliminary design should be modified to reduce technical risks and to prepare the final design.

  1. Preliminary Calculations of Bypass Flow Distribution in a Multi-Block Air Test

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Tak, Nam Il

    2011-01-01

    The development of a methodology for the bypass flow assessment in a prismatic VHTR (Very High Temperature Reactor) core has been conducted at KAERI. A preliminary estimation of variation of local bypass flow gap size between graphite blocks in the NHDD core were carried out. With the predicted gap sizes, their influence on the bypass flow distribution and the core hot spot was assessed. Due to the complexity of gap distributions, a system thermo-fluid analysis code is suggested as a tool for the core thermo-fluid analysis, the model and correlations of which should be validated. In order to generate data for validating the bypass flow analysis model, an experimental facility for a multi-block air test was constructed at Seoul National University (SNU). This study is focused on the preliminary evaluation of flow distribution in the test section to understand how the flow is distributed and to help the selection of experimental case. A commercial CFD code, ANSYS CFX is used for the analyses

  2. Mechanical design and engineering calculation of the SMCAMS magnet

    International Nuclear Information System (INIS)

    Chen Guosheng

    2001-01-01

    The basis of the mechanical design of the SMCAMS magnet, and the structure characters of the magnet and its coils are introduced. Finally, the engineering design of other parts, including deflectors, probes and accelerating electrodes are described

  3. Preliminary Design and Simulation of a Turbo Expander for Small Rated Power Organic Rankine Cycle (ORC

    Directory of Open Access Journals (Sweden)

    Roberto Capata

    2014-11-01

    , the turbine characteristics (dimensions, input and output temperature, pressure ratio, etc. have been calculated and an attempt to find the “nearly-optimal” combination has been carried out. The detailed design of a radial expander is presented and discussed. A thermo-mechanical performance study was carry out to verify structural tension and possible displacement. On the other hand, preliminary CFD analyses have been performed to verify the effectiveness of the design procedure.

  4. APPLICATION OF APM WINMACHINE SOFTWARE FOR DESIGN AND CALCULATIONS IN MECHANICAL ENGINEERING

    Directory of Open Access Journals (Sweden)

    L. O. Neduzha

    2016-04-01

    Full Text Available Purpose.To conduct the research at all stages of design, development, operation, residual operation life determination, namely, preliminary study, action principle choice, design of draft and technical projects, their optimization, preparation of design documentation and control information for automated production, comprehensive engineering analysis, it is required to use the latest computer technologies. Their use can not only present data and information in some way, but also gives the opportunity to effectively and directly interact with the information object that is created or demonstrated. Methodology.To perform engineering calculations associated with the analysis of the strength of machines, mechanisms, constructions one uses both analytical and numerical methods in practice.The most common method for analysing the stress-strain state of object models, obtaining their dynamic and stability characteristics at constant and variable modes of external load is the finite element method, which is implemented in many famous and widespread software products, providing strength calculation of models of machines, mechanisms and structures. Findings.The use of modern software for designing machine parts and various types of their joints and for strength analysis of structures is justified. Colour charts for distribution of stresses, displacement, internal efforts, safety factor and others allow accurate and quick identification of the most dangerous places in the structure. The program also provides an opportunity to «look» inside the elements and see the resulting distribution of internal force factors. Originality.The paper considered the aspects, which are unexplored at present, associated with the current state and prospects of development of industrial production, the use of software package for design and calculations in the mechanical industry. The result of the work is the justification of software application for solving problems that

  5. Modelling of extreme gusts for design calculations (NewGust)

    Energy Technology Data Exchange (ETDEWEB)

    Bierbooms, W; Cheng, Po-Wen [Delft Univ. of Technology, Inst. for Wind Energy, Delft (Netherlands); Larsen, G [Risoe National Lab., Roskilde (Denmark); Juul Pedersen, B [Vestas Wind Systems A/S, Lem (Denmark); Hansen, K [Tecnical Univ. of Denmark (Denmark)

    1999-03-01

    The main objective of the NewGust project is to come to a realistic and verified description of extreme gusts based on the stochastic properties of wind. In this paper the first results of the project are presented. Theoretical considerations indicate that the shape of extreme gusts is very sharp. Based on simulated wind time series, mean gust shapes (for several amplitudes and mean wind speeds) are determined and compared with the theoretical curves. The resemblance turned out to be very good. Furthermore, the influence of the sampling rate and the dynamics of a cup anemometer on the empirical mean gust shape are examined. The promising results are confirmed by a (preliminary) verification based on measured wind time series, available from the database on wind characteristics. The mean shape of gusts, of certain amplitude, together with their probability of occurrence can be used to obtain the distribution of the extreme response of wind turbines to gust loading. (au)

  6. The study on length and diameter ratio of nail as preliminary design for slope stabilization

    Science.gov (United States)

    Gunawan, Indra; Silmi Surjandari, Niken; Muslih Purwana, Yusep

    2017-11-01

    Soil nailing technology has been widely applied in practice for reinforced slope. The number of studies for the effective design of nail-reinforced slopes has also increased. However, most of the previous study was focused on a safety factor of the slope; the ratio of length and diameter itself has likely never been studied before. The aim of this study is to relate the length and diameter ratio of the nail with the safety factor of the 20 m height of sand slope in the various angle of friction and steepness of the slope. Simplified Bishop method was utilized to analyze the safety factor of the slope. This study is using data simulation to calculate the safety factor of the slope with soil nailing reinforcement. The results indicate that safety factor of slope stability increases with the increase of length and diameter ratio of the nail. At any angle of friction and steepness of the slope, certain effective length and diameter ratio was obtain. These results may be considered as a preliminary design for slope stabilization.

  7. A novel ultra-short scanning nuclear microprobe: Design and preliminary results

    International Nuclear Information System (INIS)

    Lebed, S.; Butz, T.; Vogt, J.; Reinert, T.; Spemann, D.; Heitmann, J.; Stachura, Z.; Lekki, J.; Potempa, A.; Styczen, J.; Sulkio-Cleff, B.

    2001-01-01

    The paper describes an optimized scanning nuclear microprobe (MP) with a new ultra-short (total length of 1.85 m) probe forming system based on a divided Russian quadruplet (DRQ) of magnetic quadrupole lenses. Modern electrostatic accelerators have a comparatively high beam brightness of about 10-25 pA/μm 2 /mrad 2 /MeV. This allows the MP proposed to provide a high lateral resolution even with large (1%) parasitic (sextupole and octupole) pole tip field components in all lenses. The features of the design permit the MP operation in the high current and low current modes with a short working distance and inexpensive quadrupole lenses. A new quadrupole doublet design has been developed for the MP. In the present work the calculated features of the new MP are compared with preliminary experimental results obtained with a similar system (total length of 2.3 m) at the INP in Cracow. The new MP is promising for studies of solids or biological samples with high resolutions (0.08-2 μm) in both modes under ambient conditions. A vertical version of the ultra-short MP can be very useful for single ion bombardments of living cells

  8. Preliminary Design Analysis of a HGD for the NHDD Program at Korea

    International Nuclear Information System (INIS)

    Song, Kee Nam; Lee, H. Y.; Lee, S. B.; Kim, Y. W.

    2007-01-01

    Korea Atomic Energy Research Institute is in the process of carrying out a Nuclear Hydrogen Development and Demonstration (NHDD) Program by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 .deg. C. A coaxial double-tube Hot Gas Duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the NHDD program. Recently, a preliminary design evaluation for the hot gas duct of the NHDD program was carried out. These preliminary design activities include a decision on the geometric dimensions, a strength evaluation, an appropriate material selection, and identifying the design code for the HGD. In this study, a preliminary strength evaluation for the HGD of the NHDD program has been undertaken based on the HTR-10 design concepts. Also, a preliminary evaluation of the creep-fatigue damage for a high temperature HGD structure has been carried out according to the draft code case for Alloy 617. Preliminary strength evaluation results for the HGD showed that the geometric dimensions of the proposed HGD would be acceptable for the design requirements

  9. Preliminary design of offshore wind turbine support structures : The importance of proper mode shape estimation

    NARCIS (Netherlands)

    Van der Male, P.

    2013-01-01

    Offshore wind turbines are highly exposed to timevarying loads. For support structures, estimation of the fatigue damage during the lifetime of the structure is an essential design aspect. This already applies for the preliminary design stage. In determining the dynamic amplification in the

  10. ICT and UD: Preliminary Study for Recommendations to Design Accessible University Courses.

    Science.gov (United States)

    Pagliara, Silvio Marcello; Sánchez Utgé, Marta; De Anna, Lucia

    2017-01-01

    Starting from the Universal Design in the educational context principles, the experiences gained during the FIRB project "Net@ccessibility" and the high-education courses for teachers' specialization on special education, this research will focus on preliminary studies in order to define the recommendations for designing accessible university courses.

  11. Status report on preliminary design activities for solar heating and cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    Information presented provides status and progress on the development of solar heating and cooling systems. The major emphasis is placed on program organization, system size definition, site identification, system approaches, heat pump and equipment design, collector procurement, and other preliminary design activities as part of the contract requirements.

  12. Preliminary neutronic design of TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    Sarikaya, B.; Tombakoglu, M.; Cecen, Y.; Kadiroglu, O. K.

    2001-01-01

    It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

  13. Preliminary calculations on the cooling rate of the Renca batholit, Sierra de San Luis, Argentina

    International Nuclear Information System (INIS)

    Lopez de Luchi, M.G.; Ostera, H.A.; Linares, E; Rosello, E.A

    2001-01-01

    Cooling rates can be used to constrain the unroofing history of plutonic-metamorphic system. Geocronological cooling rates (Spear and Parrish, 1996) can be unravelled using age calculations on minerals that were open systems and subsequently passed through their closure temperatures (Dodson, 1973) during cooling. Several age determinations on different minerals are needed in order to accurately constrain the cooling path of a pluton (Hodges 1991, Spear and Parrish, 1996 and references therein). Isotopic open-system behaviour in minerals can be modelled as volume diffusion process (Hodges, 1991 and references therein), which depends on the cooling rate of the whole system. We present the first results on the calculation of the cooling rate of the Renca batholith on the basis of the combination of both thermometric calculations and available crystallization and cooling ages (au)

  14. A Preliminary Study on Calculation of Inter-Pebble Dancoff Factor in a Pebble Type Core

    International Nuclear Information System (INIS)

    Kim, Song Hyun; Kim, Hong Chul; Kim, Soon Young; Noh, Jae Man; Kim, Jong Kyung

    2009-01-01

    The Dancoff factor is an entering probability of the neutron escaped from specific fuel kernel to another one without the interaction with moderators. Currently, Dancoff factors are mainly evaluated from stochastic methods, hence a research on analytical method is considerably insufficient in this field. In order to analytically evaluate Dancoff factor considering double-heterogeneous effect, inter-pebble and intra-pebble Dancoff factors should be calculated, respectively. Intra-pebble Dancoff factor related with the fuel kernels in one pebble was analyzed in past study. For the evaluation of inter-pebble Dancoff factor, fuel region to region Dancoff factor (FRDF) was defined and the method to calculate the FRDF is developed in this study. The result is compared with the calculation result of the MCNP5 code

  15. Preliminary design of the Carrisa Plains solar central receiver power plant. Volume III, Book 3. Appendices, Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Mouradian, E. M.

    1983-12-31

    Thermal analyses for the preliminary design phase of the Receiver of the Carrizo Plains Solar Power Plant are presented. The sodium reference operating conditions (T/sub in/ = 610/sup 0/F, T/sub out/ = 1050/sup 0/F) have been considered. Included are: Nominal flux distribution on receiver panal, Energy input to tubes, Axial temperature distribution; sodium and tubes, Sodium flow distribution, Sodium pressure drop, orifice calculations, Temperature distribution in tube cut (R-0), Backface structure, and Nonuniform sodium outlet temperature. Transient conditions and panel front face heat losses are not considered. These are to be addressed in a subsequent design phase. Also to be considered later are the design conditions as variations from the nominal reference (operating) condition. An addendum, designated Appendix C, has been included describing panel heat losses, panel temperature distribution, and tube-manifold joint thermal model.

  16. Use of the Local Variation Methods for Nuclear Design Calculations

    International Nuclear Information System (INIS)

    Zhukov, A.I.

    2006-01-01

    A new problem-solving method for steady-state equations, which describe neutron diffusion, is presented. The method bases on a variation principal for steady-state diffusion equations and direct search the minimum of a corresponding functional. Benchmark problem calculation for power of fuel assemblies show ∼ 2% relative accuracy

  17. DESIGN AND CALCULATION OF AERODROMECOAING WITH HEATED SURFACE LAYERS

    Directory of Open Access Journals (Sweden)

    Vadim G. Piskunov

    2009-04-01

    Full Text Available  The developed constructions with heated by surface layers for aerodromes and auto roads when developed composition of electroconductive concrete reinforced with chemical electrical conductive fibres being used was researched. The experimentally obtained characteristics of ended conductive concrete reinforced with fibers were presented. Calculation by developed heated construction of shell was made.

  18. Preliminary calculations of release rates from spent fuel in a tuff repository

    International Nuclear Information System (INIS)

    Apted, M.J.; O'Connell, W.J.; Lee, K.H.; MacIntyre, A.T.; Ueng, T.S.; Pigford, T.H.; Lee, W.W.L.

    1991-01-01

    Time-dependent release rates of Tc-99, I-129, Cs-135, and Np-237 have been calculated for wet-drip and moist-continuous release modes from the engineered barrier system of a potential nuclear waste repository in unsaturated tuff, representative of a possible repository at Yucca Mountain in southern Nevada. We describe the modes of water contact and of release of dissolved radionuclides to the surrounding intact rock, and the corresponding calculational models. We list the parameter values adopted, and then present numerical results, conclusions, and recommendations. 21 refs., 5 figs., 2 tabs

  19. Fort Hood Solar Total Energy Project. Volume II. Preliminary design. Part 1. System criteria and design description. Final report

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1979-01-01

    This volume documents the preliminary design developed for the Solar Total Energy System to be installed at Fort Hood, Texas. Current system, subsystem, and component designs are described and additional studies which support selection among significant design alternatives are presented. Overall system requirements which form the system design basis are presented. These include program objectives; performance and output load requirements; industrial, statutory, and regulatory standards; and site interface requirements. Material in this section will continue to be issued separately in the Systems Requirements Document and maintained current through revision throughout future phases of the project. Overall system design and detailed subsystem design descriptions are provided. Consideration of operation and maintenance is reflected in discussion of each subsystem design as well as in an integrated overall discussion. Included are the solar collector subsystem; the thermal storage subsystem, the power conversion sybsystem (including electrical generation and distribution); the heating/cooling and domestic hot water subsystems; overall instrumentation and control; and the STES building and physical plant. The design of several subsystems has progressed beyond the preliminary stage; descriptions for such subsystems are therefore provided in more detail than others to provide complete documentation of the work performed. In some cases, preliminary design parameters require specific verificaton in the definitive design phase and are identified in the text. Subsystem descriptions will continue to be issued and revised separately to maintain accuracy during future phases of the project. (WHK)

  20. The preliminary design and feasibility study of the spent fuel and high level waste repository in the Czech Republic

    International Nuclear Information System (INIS)

    Valvoda, Z.; Holub, J.; Kucerka, M.

    1996-01-01

    In the year 1993, began the Program of Development of the Spent Fuel and High Level Waste Repository in the Conditions of the Czech Republic. During the first phase, the basic concept and structure of the Program has been developed, and the basic design criteria and requirements were prepared. In the conditions of the Czech Republic, only an underground repository in deep geological formation is acceptable. Expected depth is between 500 to 1000 meters and as host rock will be granites. A preliminary variant design study was realized in 1994, that analyzed the radioactive waste and spent fuel flow from NPPs to the repository, various possibilities of transportation in accordance to the various concepts of spent fuel conditioning and transportation to the underground structures. Conditioning and encapsulation of spent fuel and/or radioactive waste is proposed on the repository site. Underground disposal structures are proposed at one underground floor. The repository will have reserve capacity for radioactive waste from NPPs decommissioning and for waste non acceptable to other repositories. Vertical disposal of unshielded canisters in boreholes and/or horizontal disposal of shielded canisters is studied. As the base term of the start up of the repository operation, the year 2035 has been established. From this date, a preliminary time schedule of the Project has been developed. A method of calculating leveled and discounted costs within the repository lifetime, for each of selected 5 variants, was used for economic calculations. Preliminary expected parametric costs of the repository are about 0,1 Kc ($0.004) per MWh, produced in the Czech NPPs. In 1995, the design and feasibility study has gone in more details to the technical concept of repository construction and proposed technologies, as well as to the operational phase of the repository. Paper will describe results of the 1995 design work and will present the program of the repository development in next period

  1. Preliminary coating design and coating developments for ATHENA

    DEFF Research Database (Denmark)

    Jakobsen, Anders Clemen; Ferreira, Desiree Della Monica; Christensen, Finn Erland

    2011-01-01

    We present initial novel coating design for ATHENA. We make use of both simple bilayer coatings of Ir and B4C and more complex constant period multilayer coatings to enhance the effective area and cover the energy range from 0.1 to 10 keV. We also present the coating technology used...... for these designs and present test results from coatings....

  2. Preliminary Calculation of the Indicators of Sustainable Development for National Radioactive Waste Management Programs

    International Nuclear Information System (INIS)

    Cheong, Jae Hak; Park, Won Jae

    2003-01-01

    As a follow up to the Agenda 21's policy statement for safe management of radioactive waste adopted at Rio Conference held in 1992, the UN invited the IAEA to develop and implement indicators of sustainable development for the management of radioactive waste. The IAEA finalized the indicators in 2002, and is planning to calculate the member states' values of indicators in connection with operation of its Net-Enabled Waste Management Database system. In this paper, the basis for introducing the indicators into the radioactive waste management was analyzed, and calculation methodology and standard assessment procedure were simply depicted. In addition, a series of innate limitations in calculation and comparison of the indicators was analyzed. According to the proposed standard procedure, the indicators for a few major countries including Korea were calculated and compared, by use of each country's radioactive waste management framework and its practices. In addition, a series of measures increasing the values of the indicators was derived so as to enhance the sustainability of domestic radioactive waste management program.

  3. Preliminary thermal/hydraulic sizing calculations for duplex tube evaporator/superheater (interchangeable units). Revision 1

    International Nuclear Information System (INIS)

    Waszink, R.P.; Hwang, J.Y.; Efferding, L.E.

    1974-06-01

    This is a preliminry thermal/hydraulic report reflecting work under Subtask 6.2 of Ref. 1.1. This report is an extension of the previous thermal/hydraulic design report. Parts of this report have been transmitted to GE. The detailed design basis, listed by source, is given. Additional details are discussed

  4. Preliminary design of a small air loop for system analysis and validation of Cathare code

    International Nuclear Information System (INIS)

    Marchand, M.; Saez, M.; Tauveron, N.; Tenchine, D.; Germain, T.; Geffraye, G.; Ruby, G.P.

    2007-01-01

    The French Atomic Energy Commission (Cea) is carrying on the design of a Small Air Loop for System Analysis (SALSA), devoted to the study of gas cooled nuclear reactors behaviour in normal and incidental/accidental operating conditions. The reduced size of the SALSA components compared to a full-scale reactor and air as gaseous coolant instead of Helium will allow an easy management of the loop. The main purpose of SALSA will be the validation of the associated thermal hydraulic safety simulation codes, like CATHARE. The main goal of this paper is to present the methodology used to define the characteristics of the loop. In a first step, the study has been focused on a direct-cycle system for the SALSA loop with few global constraints using a similarity analysis to support the definition and design of the loop. Similarity requirements have been evaluated to determine the scale factors which have to be applied to the SALSA loop components. The preliminary conceptual design of the SALSA plant with a definition of each component has then be carried out. The whole plant has been modelled using the CATHARE code. Calculations of the SALSA steady-state in nominal conditions and of different plant transients in direct-cycle have been made. The first system results obtained on the global behaviour of the loop confirm that SALSA can be representative of a Gas-Cooled nuclear reactor with some minor design modifications. In a second step, the current prospects focus on the SALSA loop capability to reproduce correctly the heat transfer occurring in specific incidental situations. Heat decay removal by natural convection is a crucial point of interest. The first results show that the behaviour and the efficiency of the loop are strongly influenced by the definition of the main parameters for each component. A complete definition of SALSA is under progress. (authors)

  5. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  6. Poster - 08: Preliminary Investigation into Collapsed-Cone based Dose Calculations for COMS Eye Plaques

    International Nuclear Information System (INIS)

    Morrison, Hali; Menon, Geetha; Sloboda, Ron

    2016-01-01

    Purpose: To investigate the accuracy of model-based dose calculations using a collapsed-cone algorithm for COMS eye plaques loaded with I-125 seeds. Methods: The Nucletron SelectSeed 130.002 I-125 seed and the 12 mm COMS eye plaque were incorporated into a research version of the Oncentra® Brachy v4.5 treatment planning system which uses the Advanced Collapsed-cone Engine (ACE) algorithm. Comparisons of TG-43 and high-accuracy ACE doses were performed for a single seed in a 30×30×30 cm 3 water box, as well as with one seed in the central slot of the 12 mm COMS eye plaque. The doses along the plaque central axis (CAX) were used to calculate the carrier correction factor, T(r), and were compared to tabulated and MCNP6 simulated doses for both the SelectSeed and IsoAid IAI-125A seeds. Results: The ACE calculated dose for the single seed in water was on average within 0.62 ± 2.2% of the TG-43 dose, with the largest differences occurring near the end-welds. The ratio of ACE to TG-43 calculated doses along the CAX (T(r)) of the 12 mm COMS plaque for the SelectSeed was on average within 3.0% of previously tabulated data, and within 2.9% of the MCNP6 simulated values. The IsoAid and SelectSeed T(r) values agreed within 0.3%. Conclusions: Initial comparisons show good agreement between ACE and MC doses for a single seed in a 12 mm COMS eye plaque; more complicated scenarios are being investigated to determine the accuracy of this calculation method.

  7. Poster - 08: Preliminary Investigation into Collapsed-Cone based Dose Calculations for COMS Eye Plaques

    Energy Technology Data Exchange (ETDEWEB)

    Morrison, Hali; Menon, Geetha; Sloboda, Ron [Cross Cancer Institute, Edmonton, AB, and University of Alberta, Edmonton, AB, Cross Cancer Institute, Edmonton, AB, and University of Alberta, Edmonton, AB, Cross Cancer Institute, Edmonton, AB, and University of Alberta, Edmonton, AB (Canada)

    2016-08-15

    Purpose: To investigate the accuracy of model-based dose calculations using a collapsed-cone algorithm for COMS eye plaques loaded with I-125 seeds. Methods: The Nucletron SelectSeed 130.002 I-125 seed and the 12 mm COMS eye plaque were incorporated into a research version of the Oncentra® Brachy v4.5 treatment planning system which uses the Advanced Collapsed-cone Engine (ACE) algorithm. Comparisons of TG-43 and high-accuracy ACE doses were performed for a single seed in a 30×30×30 cm{sup 3} water box, as well as with one seed in the central slot of the 12 mm COMS eye plaque. The doses along the plaque central axis (CAX) were used to calculate the carrier correction factor, T(r), and were compared to tabulated and MCNP6 simulated doses for both the SelectSeed and IsoAid IAI-125A seeds. Results: The ACE calculated dose for the single seed in water was on average within 0.62 ± 2.2% of the TG-43 dose, with the largest differences occurring near the end-welds. The ratio of ACE to TG-43 calculated doses along the CAX (T(r)) of the 12 mm COMS plaque for the SelectSeed was on average within 3.0% of previously tabulated data, and within 2.9% of the MCNP6 simulated values. The IsoAid and SelectSeed T(r) values agreed within 0.3%. Conclusions: Initial comparisons show good agreement between ACE and MC doses for a single seed in a 12 mm COMS eye plaque; more complicated scenarios are being investigated to determine the accuracy of this calculation method.

  8. EXPLOSION POTENTIAL ASSESSMENT OF HEAT EXCHANGER NETWORK AT THE PRELIMINARY DESIGN STAGE

    Directory of Open Access Journals (Sweden)

    MOHSIN PASHA

    2016-07-01

    Full Text Available The failure of Shell and Tube Heat Exchangers (STHE is being extensively observed in the chemical process industries. This failure can cause enormous production loss and have a potential of dangerous consequences such as an explosion, fire and toxic release scenarios. There is an urgent need for assessing the explosion potential of shell and tube heat exchanger at the preliminary design stage. In current work, inherent safety index based approach is used to resolve the highlighted issue. Inherent Safety Index for Shell and Tube Heat Exchanger (ISISTHE is a newly developed index for assessing the inherent safety level of a STHE at the preliminary design stage. This index is composed of preliminary design variables and integrated with the process design simulator (Aspen HYSYS. Process information can easily be transferred from process design simulator to MS Excel spreadsheet owing to this integration. This index could potentially facilitate the design engineer to analyse the worst heat exchanger in the heat exchanger network. Typical heat exchanger network of the steam reforming process is presented as a case study and the worst heat exchanger of this network has been identified. It is inferred from this analysis that shell and tube heat exchangers possess high operating pressure, corrected mean temperature difference (CMTD and flammability and reactive potential needs to be critically analysed at the preliminary design stage.

  9. Preliminary design analysis of the ALT-II limiter for TEXTOR

    International Nuclear Information System (INIS)

    Koski, J.A.; Boyd, R.D.; Kempka, S.M.; Romig, A.D. Jr.; Smith, M.F.; Watson, R.D.; Whitley, J.B.; Conn, R.W.; Grotz, S.P.

    1984-01-01

    Installation of a large toroidal belt pump limiter, Advanced Limiter Test II (ALT-II), on the TEXTOR tokamak at Juelich, FRG is anticipated for early 1986. This paper discusses the preliminary mechanical design and materials considerations undertaken as part of the feasibility study phase for ALT-II. Since the actively cooled limiter blade is the component in direct contact with the plasma edge, and thus subject to the severe plasma environment, most preliminary design efforts have concentrated on analysis of the blade. The screening process which led to the recommended preliminary design consisting of a dispersion strenghthened copper or OFHC copper cover plate over an austenitic stainless steel base plate is discussed. A 1 to 3 mm thick low atomic number coating consisting of a graded plasma-sprayed Silicon Carbide-Aluminium composite is recommended subject to further experiment and evaluation. Thermal-hydraulic and stress analyses of the limiter blade are also discussed. (orig.)

  10. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

    1997-01-01

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  11. Preliminary research on design of traveling wave reactor

    International Nuclear Information System (INIS)

    Yan Mingyu; Chen Bin; Feng Linna; Zhang Yong

    2015-01-01

    An engineering feasible conceptual core design of large scale (e.g. 1OOOMWe output) TWR is proposed with investigation and qualitative optimization on the proper design of fuel element structure, fuel pellet, liquid metal filling gap, fuel assembly structure, core reflector and shielding and shutdown control rods. The optimized design presents a flatten radial neutron flux with a better equivalent state distribution, which means the long term burning state could be defined by initial core design and further corrected by the travelling wave progress. The optimized fuel structure improves the flow distribution between the central, parallel and corner channels. Furthermore, the power control of TWR could be implemented by the adjusting of coolant pump rotation speed as the change of coolant flow. Though the load rejection and power control between 15% to 100% nominal power could not be fulfilled by flow control without the participation of bank A control rods. (authors)

  12. The Preliminary Design and Fabrication of a Manually Operated ...

    African Journals Online (AJOL)

    A ten (10) tonnes capacity agro waste manual briquetting machine have been designed and fabricated using locally available materials. The machine principal parts are made of frame, compaction chamber and base plate . Compaction ...

  13. Preliminary design of an osmotic-type salinity gradient energy converter. Phase I, design effort

    Energy Technology Data Exchange (ETDEWEB)

    1979-04-30

    The base case that was studied for this Phase I Interim Report is a 50 kWe design with 3.5% salt water (seawater) on one side and saturated salt water on the other side of the semi-permeable membrane. This case included a solar evaporation pond. The report includes system descriptions, system component descriptions, siting restrictions, environmental considerations, pretreatment, membrane characteristics, preliminary system capital costs, and recommendations for further work. During the course of the study and investigations, it was decided to extend the review to develop an additional basic flow sheet using brackish water instead of seawater with a solar pond. This option requires reduced flow rates and therefore can utilize smaller and less expensive components as compared to the seawater base case. Based on data for reverse osmosis water purification systems, the operating costs for pretreatment and labor would also be expected to be less for the brackish water system than for the seawater system. Finally, the use of brackish water systems greatly increases the potential number of sites available for a practical Osmo-Hydro Power System.

  14. Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 1, Report

    International Nuclear Information System (INIS)

    1986-01-01

    This design report describes the NUS Preliminary Design of the Prototype Spent Nuclear Fuel Rod Consolidation Equipment for the Department of Energy. The sections of the report elaborate on each facet of the preliminary design. A concept summary is provided to assist the reader in rapidly understanding the complete design. The NUS Prototype Spent Fuel Rod Consolidation System is an automatically controlled system to consolidate a minimum of 750 MT (heavy metal)/year of US commercial nuclear reactor fuel, at 75% availability. The system is designed with replaceable components utilizing the latest state-of-the-art technology. This approach gives the system the flexibility to be developed without costly development programs, yet accept new technology as it evolves over the next ten years. Capability is also provided in the system design to accommodate a wide variety of fuel conditions and to recover from any situation which may arise

  15. Preliminary design of reactor coolant pump canned motor for AC600

    International Nuclear Information System (INIS)

    Deng Shaowen

    1998-01-01

    The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

  16. Correction of rhodium detector signals for comparison to design calculations

    International Nuclear Information System (INIS)

    Judd, J.L.; Chang, R.Y.; Gabel, C.W.

    1989-01-01

    Rhodium detectors are used in many commercial pressurized water reactors PWRs [pressurized water reactor] as in-core neutron detectors. The signals from the detectors are the result of neutron absorption in 103 Rh and the subsequent beta decay of 104 Rh to 104 Pd. The rhodium depletes ∼1% per full-power month, so corrections are necessary to the detector signal to account for the effects of the rhodium depletion. These corrections result from the change in detector self-shielding with rhodium burnup and the change in rhodium concentration itself. Correction for the change in rhodium concentration is done by multiplication of the factor N(t)/N 0 , where N(t) is the rhodium concentration at time t and N 0 is the initial rhodium concentration. The calculation of the self-shielding factor is more complicated and is presented. A self-shielding factor based on the fraction of rhodium remaining was calculated with the CASMO-3 code. The results obtained from our comparisons of predicted and measured in-core detector signals show that the CASMO-3/SIMULATE-3 code package is an effective tool for estimating pin peaking and power distributions

  17. City of Hoboken Energy Surety Analysis: Preliminary Design Summary

    Energy Technology Data Exchange (ETDEWEB)

    Stamp, Jason Edwin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Military and Energy Systems Analysis Dept.; Baca, Michael J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Military and Energy Systems Analysis Dept.; Munoz-Ramos, Karina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Military and Energy Systems Analysis Dept.; Schenkman, Benjamin L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Military and Energy Systems Analysis Dept.; Eddy, John P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Systems Readiness and Sustainment Technology Dept.; Smith, Mark A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Systems Readiness and Sustainment Technology Dept.; Guttromson, Ross [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Electric Power Systems Research Dept.; Henry, Jordan M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Critical Infrastructure Systems Dept.; Jensen, Richard Pearson [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Geomechanics Dept.

    2014-09-01

    In 2012, Hurricane Sandy devastated much of the U.S. northeast coastal areas. Among those hardest hit was the small community of Hoboken, New Jersey, located on the banks of the Hudson River across from Manhattan. This report describes a city-wide electrical infrastructure design that uses microgrids and other infrastructure to ensure the city retains functionality should such an event occur in the future. The designs ensure that up to 55 critical buildings will retain power during blackout or flooded conditions and include analysis for microgrid architectures, performance parameters, system control, renewable energy integration, and financial opportunities (while grid connected). The results presented here are not binding and are subject to change based on input from the Hoboken stakeholders, the integrator selected to manage and implement the microgrid, or other subject matter experts during the detailed (final) phase of the design effort.

  18. Preliminary design implications of SSC fixed-target operation

    International Nuclear Information System (INIS)

    Zisman, M.S.

    1984-06-01

    This paper covers some of the accelerator physics issues relevant to a possible fixed-target operating mode for the Superconducting Super Collider (SSC). In the brief time available, no attempt has been made to design this capability into the SSC. Rather, I have tried to evaluate what the performance of such a machine might be, and to indicate the hardware implications and extraction considerations that would be part of an actual design study. Where appropriate, parameters and properties of the present LBL design for the SSC have been used; these should be taken as being representative of the general class of small-aperture, high-field colliders being considered by the accelerator physics community. Thus, the numerical examples given here must ultimately be reexamined in light of the actual parameters of the particular accelerator being considered

  19. Preliminary results on food consumption rates for off-site dose calculation of nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Gab Bock; Chung, Yang Geun; Bang, Sun Young; Kang, Duk Won

    2005-01-01

    The Internal dose by food consumption mostly account for radiological dose of public around nuclear power plants(NPP). But, food consumption rate applied to off-site dose calculation in Korea which is the result of field investigation around Kori NPP by the KAERI in 1988. is not reflected of the latest dietary characteristics. The Ministry of Health and Welfare Affairs has investigated the food and nutrition of nations every 3 years based on the Law of National Health Improvement. To update the food consumption rates of the maximum individual, the analysis of the national food investigation results and field surveys around nuclear power plant sites have been carried out

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

    International Nuclear Information System (INIS)

    Tanaka, Mitsugu

    1978-01-01

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

  1. Design review report for rotary mode core sample truck (RMCST) modifications for flammable gas tanks, preliminary design

    International Nuclear Information System (INIS)

    Corbett, J.E.

    1996-02-01

    This report documents the completion of a preliminary design review for the Rotary Mode Core Sample Truck (RMCST) modifications for flammable gas tanks. The RMCST modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to validate basic design assumptions and concepts to support a path forward leading to a final design. The conclusion reached by the review committee was that the design was acceptable and efforts should continue toward a final design review

  2. TPX: Contractor preliminary design review. Volume 2, PF systems engineering

    International Nuclear Information System (INIS)

    Calvin, H.A.

    1995-01-01

    This system development specification covers the Poloidal Field (PF) Magnet System, WBS 14 in the Princeton Plasma Physics Laboratory TPX Program to build a tokamak fusion reactor. This specification establishes the performance, design, development and test requirements of the PF Magnet System

  3. TPX: Contractor preliminary design review. Volume 2, PF systems engineering

    Energy Technology Data Exchange (ETDEWEB)

    Calvin, H.A. [Lawrence Livermore National Lab., CA (United States)

    1995-07-28

    This system development specification covers the Poloidal Field (PF) Magnet System, WBS 14 in the Princeton Plasma Physics Laboratory TPX Program to build a tokamak fusion reactor. This specification establishes the performance, design, development and test requirements of the PF Magnet System.

  4. Preliminary design considerations for the stage 1 PEP lattice

    International Nuclear Information System (INIS)

    Helm, R.H.; Lee, M.J.

    1974-07-01

    A general description of the proposed PEP e + e - storage ring is discussed in the paper. We discuss the lattice and its operating characteristics in more detail, show how the design luminosity operative regions may be met and outline the limits of the operative regions of the beam parameters in several modes of operation. 18 refs., 16 figs., 1 tab

  5. Isotope Production Facility Conceptual Thermal-Hydraulic Design Review and Scoping Calculations

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Shelton, J.D.

    1998-01-01

    The thermal-hydraulic design of the target for the Isotope Production Facility (IPF) is reviewed. In support of the technical review, scoping calculations are performed. The results of the review and scoping calculations are presented in this report

  6. Preliminary Criticality Calculation on Conceptual Deep Borehole Disposal System for Trans-metal Waste during Operational Phase

    International Nuclear Information System (INIS)

    Kim, In Young; Choi, Heui Joo; Cho, Dong Geun

    2013-01-01

    The primary function of any repository is to prevent spreading of dangerous materials into surrounding environment. In the case of high-level radioactive waste repository, radioactive material must be isolated and retarded during sufficient decay time to minimize radiation hazard to human and surrounding environment. Sub-criticality of disposal canister and whole disposal system is minimum requisite to prevent multiplication of radiation hazard. In this study, criticality of disposal canister and DBD system for trans-metal waste is calculated to check compliance of sub-criticality. Preliminary calculation on criticality of conceptual deep borehole disposal system and its canister for trans-metal waste during operational phase is conducted in this study. Calculated criticalities at every temperature are under sub-criticalities and criticalities of canister and DBD system considering temperature are expected to become 0.34932 and 0.37618 approximately. There are obvious limitations in this study. To obtain reliable data, exact elementary composition of each component, system component temperature must be specified and applied, and then proper cross section according to each component temperature must be adopted. However, many assumptions, for example simplified elementary concentration and isothermal component temperature, are adopted in this study. Improvement of these data must be conducted in the future work to progress reliability. And, post closure criticality analyses including geo, thermal, hydro, mechanical, chemical mechanism, especially fissile material re-deposition by precipitation and sorption, must be considered to ascertain criticality safety of DBD system as a future work

  7. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    International Nuclear Information System (INIS)

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

  8. The Guardian: Preliminary design of a close air support aircraft

    Science.gov (United States)

    Haag, Jonathan; Huber, David; Mcinerney, Kelly; Mulligan, Greg; Pessin, David; Seelos, Michael

    1991-01-01

    One design is presented of a Close Air Support (CAS) aircraft. It is a canard wing, twin engine, twin vertical tail aircraft that has the capability to cruise at 520 knots. The Guardian contains state of the art flight control systems. Specific highlights of the Guardian include: (1) low cost (the acquisition cost per airplane is $13.6 million for a production of 500 airplanes); (2) low maintenance (it was designed to be easily maintainable in unprepared fields); and (3) high versatility (it can perform a wide range of missions). Along with being a CAS aircraft, it is capable of long ferry missions, battlefield interdiction, maritime attack, and combat rescue. The Guardian is capable of a maximum ferry of 3800 nm, can takeoff in a distance of 1700 ft, land in a ground roll distance of 1644 ft. It has a maximum takeoff weight of 48,753 lbs, and is capable of carrying up to 19,500 lbs of ordinance.

  9. Preliminary Calculation of the EROI for the Production of Gas in Russia

    Directory of Open Access Journals (Sweden)

    Roman Nogovitsyn

    2014-09-01

    Full Text Available Russia is one of the world’s largest producers of energy resources. Production of energy resources in Russia is profitable, both economically and in terms of the energy produced (as measured by EROI. At the present time, Russian oil and gas companies have a policy of energy saving, and data on energy consumption is given in annual reports. Based on these data, we can make the EROI calculation. In 2013, the EROI for the production, transportation and processing of gas for Open joint stock company (OJSC “Gazprom” was 79:1; for OJSC “NOVATEK”, 76:1; for OJSC “Yakutsk Fuel and Energy Company (YATEC”, only for production, 116:1. Currently, the situation in the oil and gas industry has come to a point when there is a need for the introduction of an energy audit.

  10. Preliminary Design of the Liquid Lead Corrosion Test Loop

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Cha, Jae Eun; Cho, Choon Ho; Song, Tae Yung; Kim, Hee Reyoung

    2005-01-01

    Recently, Lead-Bismuth Eutectic (LBE) or Lead has newly attracted considerable attraction as a coolant to get the more inherent safety. Above all, LBE is preferred as the coolant and target material for an Accelerator-Driven System (ADS) due to its high production rate of neutrons, effective heat removal, and good radiation damage properties. But, the LBE or Lead as a coolant has a challenging problem that the LBE or Lead is more corrosive to the construction materials and fuel cladding material than the sodium because the solubility of Ni, Cr and Fe is high. After all, the LBE or Lead corrosion has been considered as an important design limit factor of ADS and Liquid Metal cooled Fast Reactors (LMFR). The Korea Atomic Energy Research Institute (KAERI) has been developing an ADS called HYPER. HYPER is designed to transmute Transuranics (TRU), Tc-99 and I-129 coming from Pressurized Water Reactors (PWRs) and uses an LBE as a coolant and target material. Also, an experimental apparatuses for the compatibility of fuel cladding and structural material with the LBE or Lead are being under the construction or design. The main objective of the present paper is introduction of Lead corrosion test loop which will be built the upside of the LBE corrosion test loop by the end of October of 2005

  11. Preliminary design of a 10 MV ion accelerator

    International Nuclear Information System (INIS)

    Fessenden, T.J.; Celata, C.M.; Faltens, A.

    1986-06-01

    At the low energy end of an induction linac HIF driver the beam current is limited by our ability to control space charge by a focusing system. As a consequence, HIF induction accelerator designs feature simultaneous acceleration of many beams in parallel within a single accelerator structure. As the speed of the beams increase, the focusing system changes from electrostatic to magnetic quadrupoles with a corresponding increase in the maximum allowable current. At that point the beams are merged thereby decreasing the cost of the subsequent accelerator structure. The LBL group is developing an experiment to study the physics of merging and of focusing ion beams. In the design, parallel beams of ions (C + , Al + , or Al ++ ) are accelerated to several MV and merged transversely. The merged beams are then further accelerated and the growth in transverse and longitudinal emittance is determined for comparison with theory. The apparatus will then be used to study the problems associated with focusing ion beams to a small spot. Details of the accelerator design and considerations of the physics of combining beams are presented

  12. Preliminary Design Requirements Document for Project W-314

    Energy Technology Data Exchange (ETDEWEB)

    MCGREW, D.L.

    2000-04-27

    This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1.

  13. Preliminary conceptual engineering design considerations for the MX machine

    International Nuclear Information System (INIS)

    Bulmer, R.H.; Calderon, M.U.; Hibbs, S.M.; Kozman, T.A.

    1975-01-01

    The mirror experiment was designed to develop the technologies necessary to make the transition from the presently small-scale physics experiments (2XIIB and BBII) to large-scale steady-state DT burning systems, such as the Fusion Engineering Research Facility (FERF) and Controlled Thermonuclear Reactors (CTR) based on plasma confinement in open magnetic geometry. The confinement parameters in the design of the present machine include a 20-kG central field with a mirror ratio of 2 to 1 and an overall BL product approximately 5 times greater than that currently available with the 2XIIB compression coils (or a mirror-to-mirror length of 3.4 m). Several types of Yin-Yang minimum parallel B parallel geometries were studied, and a ''displaced'' Yin-Yang was chosen because the center of the machine is easily accessable between the coils and between the magnet lobes. Other important design considerations include the target plasma system, the vacuum system, and the injectors. The target plasma system includes a pellet generating system used to produce a 400-μm deuterium pellet and a two-arm laser system where the laser energy is produced from a 1-kJ, 10-GW CO 2 laser at 100 ns

  14. Preliminary Design Requirements Document for Project W-314

    International Nuclear Information System (INIS)

    MCGREW, D.L.

    2000-01-01

    This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1

  15. Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

    The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

  16. Preliminary S-CO_2 Compressor Design for Micro Modular Reactor

    International Nuclear Information System (INIS)

    Lee, Jekyoung; Cho, Seong Kuk; Kim, Seong Gu; Lee, Jeong Ik

    2016-01-01

    Due to economic benefit of S-CO_2 Brayton cycle which is came from high efficiency and compactness, active research is currently conducted by various research groups and various approaches are suggested to take benefits of S-CO_2 Brayton cycle. KAIST research team also has been working on advanced concept for application of S-CO_2 Brayton cycle to nuclear system and Micro Modular Reactor (MMR) concept was suggested. The preliminary compressor design of S-CO_2 compressor for MMR system was carried out to observe feasibility of compressor design. Preliminary S-CO_2 compressor design for MMR system was successfully conducted and some issues are discovered from the design study. From the previous work done by Cho, conceptual design for MMR system was provided. Thus, further preliminary design should be carried out to obtain feasible S-CO_2 compressor design for MMR system. KAIST_TMD which is turbomachinery in-house code for real gases including S-CO_2 is continuously updated and currently it has 3D geometry construction and design optimization capability

  17. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  18. Preliminary design study of the Tandem Mirror Reactor (TMR)

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Carlson, G.A.

    1978-01-01

    This report describes work done in Fiscal Year 1977 by the Fusion Reactor Studies Group of LLL on the conceptual design of a 1000-MW(e) Tandem Mirror Reactor (TMR). The high Q (defined as the ratio of fusion power to injection power) predicted for the TMR (approximately 5) reduces the recirculating power to a nondominant problem and results in an attractive mirror fusion power plant. The fusion plasma of the TMR is contained in the 100-m-long central cell where the magnetic field strength is a modest 2 T. The blanket for neutron energy recovery and tritium breeding is cylindrical and, along with the solenoidal magnet, is divided into 3-m-long modules to facilitate maintenance. The central cell is fueled (but not heated) by the injection of low-energy neutral beams near its ends. Thus, the central cell is simple and of low technology. The end-cell plasmas must be of high density and high energy in order to plug and heat (via the electrons) the central-cell plasma. The present conceptual design uses 1.2-MeV neutral-beam injection for the end plugs and a cryogenic-aluminum, Yin-Yang magnet that produces an incremental field of about 1 T over a field of 16 T produced by a pair of Nb 3 Sn superconducting solenoids. Important design problems remain in both the neutral-beam injector and in the end-plug magnet. Also remaining are important physics questions such as alpha-beam particle transport and end-plug stability. These questions are discussed at length in the report and suggestions for future work are given

  19. Fast electrochemical membrane actuator: Design, fabrication and preliminary testing

    Science.gov (United States)

    Uvarov, I. V.; Postnikov, A. V.; Shlepakov, P. S.; Naumov, V. V.; Koroleva, O. M.; Izyumov, M. O.; Svetovoy, V. B.

    2017-11-01

    An actuator based on water electrolysis with a fast change of voltage polarity is presented. It demonstrates a new actuation principle allowing significant increase the operation frequency of the device due to fast termination of the produced gas. The actuator consists of a working chamber with metallic electrodes and supplying channels filled with an electrolyte. The chamber is formed in a layer of SU-8 and covered by a flexible polydimethylsiloxane membrane, which deforms as the pressure in the chamber increases. Design, fabrication procedure, and first tests of the actuator are described.

  20. Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 2, Drawings

    International Nuclear Information System (INIS)

    1986-01-01

    This volume consists of 65 E size drawings and 4 sketches of the NUS spent fuel rod consolidation equipment. The drawings have been grouped into categories; a detailed list of the drawings is included. The sketches prepared during the preliminary design process have been included

  1. Notification: Preliminary Research on EPA's Design for the Environment Product Labeling Program OIG

    Science.gov (United States)

    Project #OPE-FY14-4012, November 06, 2013. The Office of Inspector General (OIG) is starting preliminary research on the U.S. Environmental Protection Agency’s (EPA’s) Design for the Environment (DfE) Product Labeling Program.

  2. Preliminary design analysis of the ALT-II limiter for TEXTOR

    International Nuclear Information System (INIS)

    Koski, J.A.; Boyd, R.D.; Kempka, S.M.; Romig, A.D. Jr.; Smith, M.F.; Watson, R.D.; Whitley, J.B.; Conn, R.W.; Grotz, S.P.

    1983-01-01

    Installation of a large toroidal belt pump limiter, Advanced Limiter Test II (ALT-II), on the TEXTOR tokamak at Juelich, FRG is anticipated for early 1986. This paper discusses the preliminary mechanical design and materials considerations undertaken as part of the feasibility study phase for ALT-II

  3. Current Mooring Design in Partner WECs and Candidates for Preliminary Analysis

    DEFF Research Database (Denmark)

    Thomsen, Jonas Bjerg; Ferri, Francesco; Kofoed, Jens Peter

    This report is the combined report of Commercial Milestone "CM1: Design and Cost of Current Mooring Solutions of Partner WECs" and Milestone "M3: Mooring Solutions for Preliminary Analysis" of the EUDP project "Mooring Solutions for Large Wave Energy Converters". The report covers a description o...

  4. 4MOST: the 4-metre Multi-Object Spectroscopic Telescope project at preliminary design review

    NARCIS (Netherlands)

    de Jong, Roelof S.; Barden, Samuel C.; Bellido-Tirado, Olga; Brynnel, Joar G.; Frey, Steffen; Giannone, Domenico; Haynes, Roger; Johl, Diana; Phillips, Daniel; Schnurr, Olivier; Walcher, Jakob C.; Winkler, Roland; Ansorge, Wolfgang R.; Feltzing, Sofia; McMahon, Richard G.; Baker, Gabriella; Caillier, Patrick; Dwelly, Tom; Gaessler, Wolfgang; Iwert, Olaf; Mandel, Holger G.; Piskunov, Nikolai A.; Pragt, Johan H.; Walton, Nicholas A.; Bensby, Thomas; Bergemann, Maria; Chiappini, Cristina; Christlieb, Norbert; Cioni, Maria-Rosa L.; Driver, Simon; Finoguenov, Alexis; Helmi, Amina; Irwin, Michael J.; Kitaura, Francisco-Shu; Kneib, Jean-Paul; Liske, Jochen; Merloni, Andrea; Minchev, Ivan; Richard, Johan; Starkenburg, Else

    2016-01-01

    We present an overview of the 4MOST project at the Preliminary Design Review. 4MOST is a major new wide-field, high-multiplex spectroscopic survey facility under development for the VISTA telescope of ESO. 4MOST has a broad range of science goals ranging from Galactic Archaeology and stellar physics

  5. SUMS preliminary design and data analysis development. [shuttle upper atmosphere mass spectrometer experiment

    Science.gov (United States)

    Hinson, E. W.

    1981-01-01

    The preliminary analysis and data analysis system development for the shuttle upper atmosphere mass spectrometer (SUMS) experiment are discussed. The SUMS experiment is designed to provide free stream atmospheric density, pressure, temperature, and mean molecular weight for the high altitude, high Mach number region.

  6. Design and preliminary biomechanical analysis of artificial cervical joint complex.

    Science.gov (United States)

    Jian, Yu; Lan-Tao, Liu; Zhao, Jian-ning; Jian-ning, Zhao

    2013-06-01

    To design an artificial cervical joint complex (ACJC) prosthesis for non-fusion reconstruction after cervical subtotal corpectomy, and to evaluate the biomechanical stability, preservation of segment movements and influence on adjacent inter-vertebral movements of this prosthesis. The prosthesis was composed of three parts: the upper/lower joint head and the middle artificial vertebrae made of Cobalt-Chromium-Molybdenum (Co-Cr-Mo) alloy and polyethylene with a ball-and-socket joint design resembling the multi-axial movement in normal inter-vertebral spaces. Biomechanical tests of intact spine (control), Orion locking plate system and ACJC prosthesis were performed on formalin-fixed cervical spine specimens from 21 healthy cadavers to compare stability, range of motion (ROM) of the surgical segment and ROM of adjacent inter-vertebral spaces. As for stability of the whole lower cervical spine, there was no significant difference of flexion, extension, lateral bending and torsion between intact spine group and ACJC prosthesis group. As for segment movements, difference in flexion, lateral bending or torsion between ACJC prosthesis group and control group was not statistically significant, while ACJC prosthesis group showed an increase in extension (P inter-vertebral ROM of the ACJC prosthesis group was not statistically significant compared to that of the control group. After cervical subtotal corpectomy, reconstruction with ACJC prosthesis not only obtained instant stability, but also reserved segment motions effectively, without abnormal gain of mobility at adjacent inter-vertebral spaces.

  7. Preliminary design of a Tandem-Mirror-Next-Step facility

    International Nuclear Information System (INIS)

    Damm, C.C.; Doggett, J.N.; Bulmer, R.H.

    1980-01-01

    The Tandem-Mirror-Next-Step (TMNS) facility is designed to demonstrate the engineering feasibility of a tandem-mirror reactor. The facility is based on a deuterium-tritium (D-T) burning, tandem-mirror device with a fusion power output of 245 MW. The fusion power density in the central cell is 2.1 MW/m 3 , with a resultant neutron wall loading of 0.5 MW/m 2 . Overall machine length is 116 m, and the effective central-cell length is 50.9 m. The magnet system includes end cells with yin-yang magnets to provide magnetohydrodynamic (MHD) stability and thermal-barrier cells to help achieve a plasma Q of 4.7 (where Q = fusion power/injected power). Neutral beams at energies up to 200 keV are used for plasma heating, fueling, and barrier pumping. Electron cyclotron resonant heating at 50 and 100 GHz is used to control the electron temperature in the barriers. Based on the resulting engineering design, the overall cost of the facility is estimated to be just under $1 billion. Unresolved physics issues include central-cell β-limits against MHD ballooning modes (the assumed reference value of β exceeds the current theory-derived limit), and the removal of thermalized α-particles from the plasma

  8. Kemper County IGCC (tm) Project Preliminary Public Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Matt; Rush, Randall; Madden, Diane; Pinkston, Tim; Lunsford, Landon

    2012-07-01

    The Kemper County IGCC Project is an advanced coal technology project that is being developed by Mississippi Power Company (MPC). The project is a lignite-fueled 2-on-1 Integrated Gasification Combined-Cycle (IGCC) facility incorporating the air-blown Transport Integrated Gasification (TRIG™) technology jointly developed by Southern Company; Kellogg, Brown, and Root (KBR); and the United States Department of Energy (DOE) at the Power Systems Development Facility (PSDF) in Wilsonville, Alabama. The estimated nameplate capacity of the plant will be 830 MW with a peak net output capability of 582 MW. As a result of advanced emissions control equipment, the facility will produce marketable byproducts of ammonia, sulfuric acid, and carbon dioxide. 65 percent of the carbon dioxide (CO{sub 2}) will be captured and used for enhanced oil recovery (EOR), making the Kemper County facility’s carbon emissions comparable to those of a natural-gas-fired combined cycle power plant. The commercial operation date (COD) of the Kemper County IGCC plant will be May 2014. This report describes the basic design and function of the plant as determined at the end of the Front End Engineering Design (FEED) phase of the project.

  9. Integrated Box Interrogation System (IBIS) Preliminary Design Study

    Energy Technology Data Exchange (ETDEWEB)

    DR. Stephen Croft; Mr. David Martancik; Dr. Brian Young; Dr. Patrick MJ Chard; Dr. Robert J Estop; Sheila Melton; Gaetano J. Arnone

    2003-01-13

    Canberra Industries has won the tendered solicitation, INEEL/EST-99-00121 for boxed waste Nondestructive Assay Development and Demonstration. Canberra will provide the Integrated Box Interrogation System (IBIS) which is a suite of assay instrumentation and a data reduction system that addresses the measurement needs for Boxed Wastes identified in the solicitation and facilitates the associated experimental program and demonstration of system capability. The IBIS system will consist of the next generation CWAM system, i.e. CWAM II, which is a Scanning Passive/Active Neutron interrogation system which we will call a Box Segmented Neutron Scanner (BSNS), combined with a physically separate Box Segmented Gamma-ray Scanning (BSGS) system. These systems are based on existing hardware designs but will be tailored to the large sample size and enhanced to allow the program to evaluate the following measurement criteria:Characterization and correction for matrix heterogeneity Characterization of non-uniform radio-nuclide and isotopic compositions Assay of high density matrices (both high-Z and high moderator contents)Correction for radioactive material physical form - such as self shielding or multiplication effects due to large accumulations of radioactive materials.Calibration with a minimal set of reference standards and representative matrices.THis document summarizes the conceptual design parameters of the IBIS and indicates areas key to the success of the project where development is to be centered. The work presented here is a collaborative effort between scientific staff within Canberra and within the NIS-6 group at LANL.

  10. Integrated Box Interrogation System (IBIS) Preliminary Design Study

    International Nuclear Information System (INIS)

    Croft, Stephen; Martancik, David; Young, Brian; Chard MJ, Patrick; Estop J, Robert; Sheila Melton; Arnone, Gaetano J.

    2003-01-01

    Canberra Industries has won the tendered solicitation, INEEL/EST-99-00121 for boxed waste Nondestructive Assay Development and Demonstration. Canberra will provide the Integrated Box Interrogation System (IBIS) which is a suite of assay instrumentation and a data reduction system that addresses the measurement needs for Boxed Wastes identified in the solicitation and facilitates the associated experimental program and demonstration of system capability. The IBIS system will consist of the next generation CWAM system, i.e. CWAM II, which is a Scanning Passive/Active Neutron interrogation system which we will call a Box Segmented Neutron Scanner (BSNS), combined with a physically separate Box Segmented Gamma-ray Scanning (BSGS) system. These systems are based on existing hardware designs but will be tailored to the large sample size and enhanced to allow the program to evaluate the following measurement criteria:Characterization and correction for matrix heterogeneity Characterization of non-uniform radio-nuclide and isotopic compositions Assay of high density matrices (both high-Z and high moderator contents)Correction for radioactive material physical form - such as self shielding or multiplication effects due to large accumulations of radioactive materials.Calibration with a minimal set of reference standards and representative matrices.THis document summarizes the conceptual design parameters of the IBIS and indicates areas key to the success of the project where development is to be centered. The work presented here is a collaborative effort between scientific staff within Canberra and within the NIS-6 group at LANL

  11. Design and preliminary assessment of Vanderbilt hand exoskeleton.

    Science.gov (United States)

    Gasser, Benjamin W; Bennett, Daniel A; Durrough, Christina M; Goldfarb, Michael

    2017-07-01

    This paper presents the design of a hand exoskeleton intended to enable or facilitate bimanual activities of daily living (ADLs) for individuals with chronic upper extremity hemiparesis resulting from stroke. The paper describes design of the battery-powered, self-contained exoskeleton and presents the results of initial testing with a single subject with hemiparesis from stroke. Specifically, an experiment was conducted requiring the subject to repeatedly remove the lid from a water bottle both with and without the hand exoskeleton. The relative times required to remove the lid from the bottles was considerably lower when using the exoskeleton. Specifically, the average amount of time required to grasp the bottle with the paretic hand without the exoskeleton was 25.9 s, with a standard deviation of 33.5 s, while the corresponding average amount of time required to grasp the bottle with the exoskeleton was 5.1 s, with a standard deviation of 1.9 s. Thus, the task time involving the paretic hand was reduced by a factor of five, while the standard deviation was reduced by a factor of 16.

  12. Calculations to support design of a nuclear material tracking system

    International Nuclear Information System (INIS)

    Carter, L.L.; Eggers, R.F.; Williams, T.L.

    1991-01-01

    The Westinghouse Hanford Company is developing a nuclear material tracking system called NTRAK for the US Department of Energy at the Savannah River site. The NTRAK system is designed to determine the position and approximate magnitude of packages of special nuclear material (SNM) moving through a nuclear plant. The NTRAK accomplishes this by using special assemblies of detectors called modules to measure the gamma radiation emitted by the SNM. After measurement, raw data are processed to determine the direction to and position of the gamma-ray source. In order for the NTRAK method of SNM tracking to work, the gamma-ray signal at the detector modules must be at least four standard deviations above background. This paper addresses the use of the Monte Carlo computer code for neutron and photon transport (MCNP) to (a) predict the radiation emitted by plutonium oxide sources and (b) predict the counting rate of NaI detectors measuring those sources

  13. External radiotherapy in macular degeneration: Our technique, dosimetric calculation, and preliminary results

    International Nuclear Information System (INIS)

    Akmansu, M.; Dirican, Bahar; Oeztuerk, Berrin; Egehan, Ibrahim; Subasi, Mahmut; Or, Meral

    1998-01-01

    Purpose: This study was performed to determine the toxicity and efficacy of external-beam radiotherapy in patients with age-related subfoveal neovascularization. Methods and Materials: Between January 1996 and September 1996, 25 patients with a mean age of 70.5 (60-84) years were enrolled. All patients underwent fluorescein angiographic evaluation and documentation of their neovascular disease prior to irradiation. A total of 25 patients were treated with a total dose of 12 Gy in 6 fractions over 8 days. We used a lens-sparing technique and patients were treated with a single lateral 6-MV photon beam. To assess the risk of radiation carcinogenesis after treatment of age-related subfoveal neovascularization, we estimated the effective dose for a standard patient on the basis of tissue-weighting factors as defined by the International Commission on Radiological Protection (ICRP). The calculations were made with TLD on a male randophantom. The lens dose was found to be 0.217 Gy per fraction. Results: No significant acute morbidity was noted. Visual acuity was maintained or improved in 76% and 80% of treated patients at their 1- and 3-month follow-up examinations, respectively. On angiographic imaging, there was stabilization of subfoveal neovascular membranes in 23 patients (92%) at 3 months after irradiation. Conclusion: Our observations on these 25 patients in this study indicate that many patients will have improved or stable vision after radiotherapy treatment with low-dose irradiation

  14. Preliminary design analysis of hot gas ducts and a intermediate heat exchanger for the nuclear hydrogen reactor

    International Nuclear Information System (INIS)

    Song, K. N.; Kim, Y. W.

    2008-01-01

    Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 .deg. C. Primary and secondary hot gas ducts with coaxial double tubes and are key components connecting a reactor pressure vessel and a intermediate heat exchanger for the nuclear hydrogen system. In this study, preliminary design analyses on the hot gas ducts and the intermediate heat exchanger were carried out. These preliminary design activities include a preliminary design on the geometric dimensions, a preliminary strength evaluation, thermal sizing, and an appropriate material selection

  15. Preliminary Axial Flow Turbine Design and Off-Design Performance Analysis Methods for Rotary Wing Aircraft Engines. Part 1; Validation

    Science.gov (United States)

    Chen, Shu-cheng, S.

    2009-01-01

    For the preliminary design and the off-design performance analysis of axial flow turbines, a pair of intermediate level-of-fidelity computer codes, TD2-2 (design; reference 1) and AXOD (off-design; reference 2), are being evaluated for use in turbine design and performance prediction of the modern high performance aircraft engines. TD2-2 employs a streamline curvature method for design, while AXOD approaches the flow analysis with an equal radius-height domain decomposition strategy. Both methods resolve only the flows in the annulus region while modeling the impact introduced by the blade rows. The mathematical formulations and derivations involved in both methods are documented in references 3, 4 for TD2-2) and in reference 5 (for AXOD). The focus of this paper is to discuss the fundamental issues of applicability and compatibility of the two codes as a pair of companion pieces, to perform preliminary design and off-design analysis for modern aircraft engine turbines. Two validation cases for the design and the off-design prediction using TD2-2 and AXOD conducted on two existing high efficiency turbines, developed and tested in the NASA/GE Energy Efficient Engine (GE-E3) Program, the High Pressure Turbine (HPT; two stages, air cooled) and the Low Pressure Turbine (LPT; five stages, un-cooled), are provided in support of the analysis and discussion presented in this paper.

  16. Integrated Box Interrogation System (IBIS) Preliminary Design Study

    CERN Document Server

    Croft, S; Chard-Mj, P; Estop, J R; Martancik, D; Sheila-Melton; Young, B

    2003-01-01

    Canberra Industries has won the tendered solicitation, INEEL/EST-99-00121 for boxed waste Nondestructive Assay Development and Demonstration. Canberra will provide the Integrated Box Interrogation System (IBIS) which is a suite of assay instrumentation and a data reduction system that addresses the measurement needs for Boxed Wastes identified in the solicitation and facilitates the associated experimental program and demonstration of system capability. The IBIS system will consist of the next generation CWAM system, i.e. CWAM II, which is a Scanning Passive/Active Neutron interrogation system which we will call a Box Segmented Neutron Scanner (BSNS), combined with a physically separate Box Segmented Gamma-ray Scanning (BSGS) system. These systems are based on existing hardware designs but will be tailored to the large sample size and enhanced to allow the program to evaluate the following measurement criteria:Characterization and correction for matrix heterogeneity Characterization of non-uniform radio-nucli...

  17. Preliminary design considerations for automatic refueling at N Reactor

    International Nuclear Information System (INIS)

    Quapp, W.J.; Yount, J.A.

    1985-01-01

    The Refueling Enhancement Program is an effort to upgrade and improve the N Reactor refueling operation. Primary goals of this effort are to reduce personnel exposure, reduce effluents to the environment, and, where possible, increase the refueling rate. Recent advances in available commercial robotics systems have prompted a look at automating the Charge/Discharge (C/D) operations. Current efforts will culminate in a conceptual design report (CDR) and accompanying economic and risk analysis in January 1986. Based on the results in that report, DOE will review the viability of the approach as a future capital project. Implementation of automation in existing plants raises questions regarding both the programmatic (how does one implement such an effort) and technical (what equipment is available; how will it be applied) concerns. This paper addresses both aspects

  18. Preliminary Design Study for a National Digital Seismograph Network

    Science.gov (United States)

    Peterson, Jon; Hutt, Charles R.

    1981-01-01

    Introduction Recently, the National Research Council published a report by the Panel on National, Regional, and Local Seismograph Networks of the Committee on Seismology in which the principal recommendation was for the establishment of a national digital seismograph network (NDSN). The Panel Report (Bolt, 1980) addresses both the need and the scientific requirements for the new national network. The purpose of this study has been to translate the scientific requirements into an instrumentation concept for the NSDS. There are literally hundreds, perhaps thousands, of seismographs in operation within the United States. Each serves an important purpose, but most have limited objectives in time, in region, or in the types of data that are being recorded. The concept of a national network, funded and operated by the Federal Government, is based on broader objectives that include continuity of time, uniform coverage, standardization of data format and instruments, and widespread use of the data for a variety of research purposes. A national digital seismograph network will be an important data resource for many years to come; hence, its design is likely to be of interest to most seismologists. Seismologists have traditionally been involved in the development and field operation of seismic systems and thus have been familiar with both the potential value and the limitations of the data. However, in recent years of increasing technological sophistication, the development of data sstems has fallen more to system engineers, and this trend is likely to continue. One danger in this is that the engineers may misinterpret scientific objectives or subordinate them to purely technological considerations. Another risk is that the data users may misuse or misinterpret the data because they are not aware of the limitations of the data system. Perhaps the most important purpose of a design study such as this is to stimulate a dialogue between system engineers and potential data users

  19. Preliminary electrostatic and mechanical design of a SINGAP-MAMuG compatible accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Grando, L. [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)], E-mail: luca.grando@igi.cnr.it; Dal Bello, S.; De Lorenzi, A. [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy); Pilan, N. [DIE, Universita di Padova, Via Gradenigo 6A, I-35100 Padova (Italy); Rizzolo, A.; Zaccaria, P. [Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)

    2009-06-15

    Each ITER NB injector shall provide 16.5 MW auxiliary power by accelerating a deuterium beam across a voltage of -1 MV. At present two possible alternatives for the accelerator are considered: the reference design, based on MAMuG electrostatic accelerator, where the total voltage is graded using five grids at intermediate steps of 200 kV, and the alternative concept, the SINGAP accelerator, for which the total voltage is held by one single gap. This paper focuses a preliminary feasibility study of integration of SINGAP accelerator grids into the support structure of a MAMuG type accelerator; the review or design of new electrostatic shields to improve the voltage withstanding capability of the system and the preliminary design of electrical and hydraulic connections routing from the bushing to the accelerator are also discussed. Electrostatic and mechanical analyses carried out to support the design are described in detail.

  20. Status of Preliminary Design on the Assembly Tools for ITER Tokamak Machine

    International Nuclear Information System (INIS)

    Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin; Moon, Jae Hwan; Kim, Byung Seok; Lee, Jae Hyuk; Shaw, Robert

    2012-01-01

    The ITER Tokamak device is principally composed of nine 40 .deg. sectors. Each 40 .deg. sector is made up of one 40 .deg. vacuum vessel (VV), two 20 .deg. toroidal filed coils (TFC) and associated vacuum vessel thermal shield (VVTS) segments which consist of one inboard and two outboard vacuum vessel thermal shields. Based on the design description document and final report prepared by the ITER organization (IO) and conceptual design, Korea has carried out the preliminary design of these assembly tools. The assembly strategy and relevant tools for the 40 .deg. sector sub-assembly and sector assembly at in-pit should be developed to satisfy the basic assembly requirements of the ITER Tokamak machine. Assembly strategy, preliminary design of the sector sub-assembly and assembly tools are described in this paper

  1. Preliminary design of the cold neutron source for the Centro Atomico Bariloche Electron LINAC Facility. I. Solid benzene as moderating material

    International Nuclear Information System (INIS)

    Torres, Lourdes; Granada, Jose R.

    2004-01-01

    We present the results of preliminary calculations performed with the code MCNP-4C relative to the neutron field behavior within the moderator for the CAB-LINAC cold neutron source, using benzene at 89 K as moderating material. Throughout the design calculations nuclear data libraries previously generated and validated were used. The optimum dimensions for a slab and a grid moderator were calculated, with and without a pre moderator, from the point of view of neutron production and the time-width of the neutron pulse. (author)

  2. Calculation and design of steel bearing structure for wind turbine

    Directory of Open Access Journals (Sweden)

    Bešević Miroslav

    2014-01-01

    Full Text Available Wind represents directed movement of the air and is caused by differences in atmospheric pressure which are caused by uneven heating of air masses. Global and local winds can be distinguished. Global winds have high altitude, while local winds occur in the ground layer of the atmosphere. Given that the global wings have high altitude they cannot be used as propellant for wind generators, but they should be known for their effects on the winds in the lower atmosphere. Modern wind turbines are made with a horizontal axle that has a system for the swiveling axis in the horizontal plane for tracking wind direction changes. They can have different number of blades, but for larger forces three blades are commonly used because they provide the greatest efficiency. Rotor diameter of these turbines depends on the strength and it ranges from 30 m for the power of 300 kW to 115 m for the power of 5 MW. Wind turbines are mounted on vertical steel tower which can be high even more than 100 m. Depending on the diameter of the turbine rotor, column is usually built as steel conical and less often as a steel-frame. This study includes analysis and design of steel tower for wind generator made by manufacturer Vestas, type V112 3MW HH 119 (power 3.2 MW for the construction of wind farm 'Kovačica'.

  3. Preliminary design of the Carrisa Plains solar central receiver power plant. Volume II. Plant specifications

    Energy Technology Data Exchange (ETDEWEB)

    Price, R. E.

    1983-12-31

    The specifications and design criteria for all plant systems and subsystems used in developing the preliminary design of Carrisa Plains 30-MWe Solar Plant are contained in this volume. The specifications have been organized according to plant systems and levels. The levels are arranged in tiers. Starting at the top tier and proceeding down, the specification levels are the plant, system, subsystem, components, and fabrication. A tab number, listed in the index, has been assigned each document to facilitate document location.

  4. Muon-catalyzed fusion experiment target and detector system. Preliminary design report

    International Nuclear Information System (INIS)

    Jones, S.E.; Watts, K.D.; Caffrey, A.J.; Walter, J.B.

    1982-03-01

    We present detailed plans for the target and particle detector systems for the muon-catalyzed fusion experiment. Requirements imposed on the target vessel by experimental conditions and safety considerations are delineated. Preliminary designs for the target vessel capsule and secondary containment vessel have been developed which meet these requirements. In addition, the particle detection system is outlined, including associated fast electronics and on-line data acquisition. Computer programs developed to study the target and detector system designs are described

  5. Mont-Terri heater test: design and preliminary results

    International Nuclear Information System (INIS)

    Garcia-Sineriz, J.L.; Fuentes, J.L.; Mayor, J.C.; Huertas, F.

    2003-01-01

    Safety and long-term behaviour of underground permanent repositories depend on a combination of several engineered and geological barriers. The properties of the geological barriers are the natural conditions of the formation, while the performance of the engineered barriers is a result of their design and construction. The properties of the engineered barriers are deeply influenced by the interactions between both geological and engineered barriers in response to the conditions expected in a high level waste repository. These interactions need to be identified and fully understood to allow their input in models describing the behaviour of the near field to predict reliably the long-term performance and safety of a repository. The Heating Experiment (HE) project, which is taking place at the Mont-Terri underground laboratory in Switzerland, is conceived as a research project to learn more about the coupled thermo-hydro-mechanical processes in a clay formation around a heat source similar to those in a potential repository, with special emphasis on the interaction between the clay host rock and the bentonite buffer that is part of the engineered barrier, under saturated conditions. This project is co-funded by the European Commission and performed as part of the fifth EURATOM framework programme, key action Nuclear Fission (1998-2002). For that purpose, a central vertical borehole of 300 mm diameter and 7 m deep was drilled and an electrical heater surrounded with a Spanish bentonite buffer was installed inside. More than seventeen boreholes were instrumented for measuring parameters such as temperatures, total pressures, radial displacements, gas/water release and for performing geo-electric tomography. A total of 112 instruments were installed. The complexity of the issues involved requires a multi-partner approach and there exists a mutual interest of national research organisations to co-operate on a European level: two national agencies, which are responsible

  6. Design and preliminary results of the IMA plasma focus experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soliman, H M; Masoud, M M [Arab Republic of Egypt, Atomic Energy Authority Plasma physics and Nuclear Fusion department, Cairo (Egypt)

    1994-12-31

    The present paper describes the design, operation and characteristics of aton 1MA plasma focus device, which built in egypt at the plasma physics department, N.R.C., atomic energy authority. The main parts of the system are: the coaxial electrodes of mather type, the expansion chamber, the condenser bank of 75 kJ stored energy, the pressurized spark gap switches and Blumlein trigger system. Measurement of the breakdown voltage between plasma focus electrodes and discharge current, using half of the condenser bank, showed that, for U{sub c} h = 32 kV, the discharge current was 0.5 Ma. In the discharge current and voltage traces a sharp drop in discharge current correspondingly to a sudden rise in voltage have been observed, which characterize the focus regime. Time structure of the x-ray emission measurements have been performed by means of scintillation detectors. by using a hydrogen gas the results showed that, the x-ray intensity is increased with increasing the hydrogen gas pressure. plasma sheath current density, J-Z distribution in axial direction during the acceleration phase of the discharge is studied, using a miniature Rogovsky coil. The results cleared that J{sub z} is increased with the axial distance from breech to muzzle at different hydrogen gas pressures. 12 figs.

  7. Design and preliminary results of the IMA plasma focus experiment

    International Nuclear Information System (INIS)

    Soliman, H.M.; Masoud, M.M.

    1993-01-01

    The present paper describes the design, operation and characteristics of aton 1MA plasma focus device, which built in egypt at the plasma physics department, N.R.C., atomic energy authority. The main parts of the system are: the coaxial electrodes of mather type, the expansion chamber, the condenser bank of 75 kJ stored energy, the pressurized spark gap switches and Blumlein trigger system. Measurement of the breakdown voltage between plasma focus electrodes and discharge current, using half of the condenser bank, showed that, for U c h = 32 kV, the discharge current was 0.5 Ma. In the discharge current and voltage traces a sharp drop in discharge current correspondingly to a sudden rise in voltage have been observed, which characterize the focus regime. Time structure of the x-ray emission measurements have been performed by means of scintillation detectors. by using a hydrogen gas the results showed that, the x-ray intensity is increased with increasing the hydrogen gas pressure. plasma sheath current density, J-Z distribution in axial direction during the acceleration phase of the discharge is studied, using a miniature Rogovsky coil. The results cleared that J z is increased with the axial distance from breech to muzzle at different hydrogen gas pressures. 12 figs

  8. A preliminary plant design study for the production of diesel from coal via fischer-tropsch synthesis

    International Nuclear Information System (INIS)

    Kamil, M.; Saleem, M.

    2010-01-01

    Pakistan's reliance on conventional means of producing energy has proven to be an inadequate strategy for overcoming it. The situation direly demands diversification of our energy resources not only to overcome current fiasco but also in planning for future. Among the other alternative sources, coal is the main source for producing cheaper electricity being available as huge reserves. This paper presents the preliminary plant design and cost estimation for the production of diesel from coal via coal gasification and fischer-Tropschs synthesis. Prelimnary design calculations and cost estimation are presented along with underlying assumptions. The results reveal that the diesel produced from this process might be cheaper than the crude oil based diesel. (author)

  9. Preliminary design characteristics of the RB fast-thermal core 'HERBE'

    International Nuclear Information System (INIS)

    Pesic, M.; Marinkovic, P.

    1989-01-01

    The 'RB' is zero power heavy water critical assembly designed in 1958 in Yugoslavia. The reactor operated using natural metal uranium, 2% enriched metal uranium, and 80% enriched UO 2 fuel of Soviet origin. A study of design of fast neutron fields began in 1976 and three fast neutron fields were designed up to 1983: the external neutron converter, the experimental fuel channel and the internal neutron converter, as the first step to fast-thermal coupled system. The preliminary design characteristics of the HERBE - a new fast - thermal core at the RB reactor are shown in this paper. (author)

  10. A study on the development plan and preliminary design of proton accelerator for nuclear application

    Energy Technology Data Exchange (ETDEWEB)

    Eom, Tae Yoon; Choi, B H; Park, C K; Chung, K S. and others

    1997-11-01

    A study on the development plan and preliminary design for the realisation of high current proton accelerator to be used as an essential component for the R and D of accelerator-driven system (ADS) for energy production and transmutation of long-lived radionuclides. Various fields of application of the accelerator such as basic nuclear physics, material science, biology, high energy physics, medicine, etc. were also investigated. From the preliminary design study, 1 GeV (20 mA) - Linac is required for the purposed of transmutation and energy production. Specification of injector, RFQ, CCTL and SL was also suggested. For the case study, a duoplasmatron ion source was designed by KAERI and fabricated by a domestic manufacturer, and the performance was also tested. (author). 71 refs., 61 tabs., 131 figs

  11. Preliminary study of an expert system for mechanical design of a pressure vessel

    International Nuclear Information System (INIS)

    Kasmuri, N.H.; Md Som, A.

    2006-01-01

    This paper describes a preliminary study of an expert system for mechanical design of a pressure vessel. The system supports the framework for the conceptual mechanical design from the initial stages within the design procedures. ASME Boiler and Pressure Vessel Code Section VIII Division 1 were applied as a design rule. The proposed methodology facilitates the development of knowledge base acquisition, knowledge base construction and the prototype implementation. This study characterizes a knowledge base (procedure) of mechanical design of a pressure vessel subjected to internal pressure including all design parameters; i.e. temperature, shell thickness, selection of materials of constructions, stress analysis procedure, support and ancillary items. The rationalization of the mechanical design is shown in the form of a schematic flow diagram. A Kappa PC expert system shell is used as a tool to develop the prototype software. It provides graphical representation for creating objects, hierarchies and rules for knowledge base used in pressure vessel design. (Author)

  12. Preliminary design report: Prototypical Spent Fuel Consolidation Equipment Demonstration Project: Phase 1

    International Nuclear Information System (INIS)

    Blissell, W.H.; Ciez, A.P.; Mitchell, J.L.; Winkler, C.J.

    1986-12-01

    This document describes the Westinghouse Preliminary Design for the Prototypical Consolidation Demonstration Project per Department of Energy (DOE) Contract No. DE-AC07-86ID12649 and under direction of the DOE Idaho Operations Office. The preliminary design is the first step to providing the Department of Energy with a fully qualified, licensable, cost-effective spent fuel rod consolidation system. The design was developed using proven technologies and equipment to create an innovative approach to previous rod consolidation concepts. These innovations will better enable the Westinghouse system to: consolidate fuel rods in a precise, fully-controlled, accountable manner; package all rods from two PWR fuel assemblies or from four BWR fuel assemblies in one 8.5 inch square consolidated rods canister; meet all functional requirements; operate with all fuel types common to the US commercial nuclear industry with minimal tooling changeouts; and meet consolidation production process rates, while maintaining operator and public health and safety. This Preliminary Design Report provides both detailed descriptions of the equipment required to perform the rod consolidation process and the supporting analyses to validate the design

  13. Ocean thermal energy conversion (OTEC) power system development. Preliminary design report, Appendices, Part 1 (Final)

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-04

    The objective of this project is the development of a preliminary design for a full-sized, closed cycle, ammonia power system module for the 100 MWe OTEC demonstration plant. In turn, this demonstration plant is to demonstrate, by 1984, the operation and performance of an Ocean Thermal Power Plant having sufficiently advanced heat exchanger design to project economic viability for commercial utilization in the late 1980's and beyond. Included in this power system development are the preliminary designs for a proof-of-concept pilot plant and test article heat exchangers which are scaled in such a manner as to support a logically sequential, relatively low-cost development of the full-scale power system module. The conceptual designs are presented for the demonstration plant power module, the proof-of-concept pilot plant, and for a pair of test article heat exchangers. Costs associated with the design, development, fabrication, checkout, delivery, installation, and operation are included. The accompanying design and producibilty studies on the full-scale power system module project the performance/economics for the commercial plant. This section of the report contains appendices on the developed computer models, water system dynamic studies, miscellaneous performance analysis, materials and processes, detailed equipment lists, turbine design studies, tube cleaner design, ammonia leak detection, and heat exchanger design supporting data. (WHK)

  14. Preliminary design for hot dirty-gas control-valve test facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    This report presents the results of a preliminary design and cost estimating effort for a facility for the testing of control valves in Hot Dirty Gas (HDGCV) service. This design was performed by Mittelhauser Corporation for the United States Department of Energy's Morgantown Energy Technology Center (METC). The objective of this effort was to provide METC with a feasible preliminary design for a test facility which could be used to evaluate valve designs under simulated service conditions and provide a technology data base for DOE and industry. In addition to the actual preliminary design of the test facility, final design/construction/operating schedules and a facility cost estimate were prepared to provide METC sufficient information with which to evaluate this design. The bases, assumptions, and limitations of this study effort are given. The tasks carried out were as follows: METC Facility Review, Environmental Control Study, Gas Generation Study, Metallurgy Review, Safety Review, Facility Process Design, Facility Conceptual Layout, Instrumentation Design, Cost Estimates, and Schedules. The report provides information regarding the methods of approach used in the various tasks involved in the completion of this study. Section 5.0 of this report presents the results of the study effort. The results obtained from the above-defined tasks are described briefly. The turnkey cost of the test facility is estimated to be $9,774,700 in fourth quarter 1979 dollars, and the annual operating cost is estimated to be $960,000 plus utilities costs which are not included because unit costs per utility were not available from METC.

  15. Preliminary design of bellows for the DNB beam source by EJMA and FE linear analysis

    International Nuclear Information System (INIS)

    Trapasiya, Shobhit; Muvvala, Venkata Nagaraju; Rambilas, P.; Gangadharan, Roopesh; Rotti, Chandramouli; Chakraborty, Arun Kumar; Sharma, Dheeraj Kumar

    2015-01-01

    In piping system, U-shaped Bellows are widely used among flexible elements. In general, bellows are typically design for Fatigue behavior according to the EJMA standard based on empirically generated fatigue curves. The present work proposes a methodology in the design of bellows by design by analyses and validates its design by EJMA standard. A linear FE approach is chosen to in line with the EJMA standard. The proposed methodology is benchmarked with the available literatures. The same practice is implemented in the preliminary design of a U-shaped bellows in the water line circuits of DNB beam source. DNB Beam Source is a negative ion source-based neutral beam generator for ITER operates at 100KV. The beam divergence (intrinsic) and magnetic fields from ITER torus causes deflection of beams. This calls for beam optic alignment, which are assured by BS Movement mechanism system. To accomplish the above movement requirements, bellows, which is a stringent of its kind (± 22 mm axial, ± 45 mm lateral within 400mm available space with single ply), is designed between the beam source and possible rigid interface-cooling lines coming from HVB. The paper describes right from conceptual stage to preliminary design. Optimization tools are adopted in the selecting bellow dimensions using MATLAB. At the end a coordinated approach between FE based assessment (in ANSYS) and widely applied code, EJMA is implemented for the validation of design and found FE approach is a very conservative than later in the present case. (author)

  16. Preliminary Design and Analysis of an In-plane PRSEUS Joint

    Science.gov (United States)

    Lovejoy, Andrew E.; Poplawski, Steven

    2013-01-01

    As part of the National Aeronautics and Space Administration's (NASA's) Environmentally Responsible Aviation (ERA) program, the Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) has been designed, developed and tested. However, PRSEUS development efforts to date have only addressed joints required to transfer bending moments between PRSEUS panels. Development of in-plane joints for the PRSEUS concept is necessary to facilitate in-plane transfer of load from PRSEUS panels to an adjacent structure, such as from a wing panel into a fuselage. This paper presents preliminary design and analysis of an in-plane PRSEUS joint for connecting PRSEUS panels at the termination of the rod-stiffened stringers. Design requirements are provided, the PRSEUS blade joint concept is presented, and preliminary design changes and analyses are carried out to examine the feasibility of the proposed in-plane PRSEUS blade joint. The study conducted herein focuses mainly on the PRSEUS structure on one side of the joint. In particular, the design requirements for the rod shear stress and bolt bearing stress are examined. A PRSEUS blade joint design was developed that demonstrates the feasibility of this in-plane PRSEUS joint concept to terminate the rod-stiffened stringers. The presented design only demonstrates feasibility, therefore, some areas of refinement are presented that would lead to a more optimum and realistic design.

  17. Ocean thermal energy conversion (OTEC). Power system development. Preliminary design report, final

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-04

    The preliminary design of the 10 MWe OTEC power module and the 200 kWe test articles is given in detail. System operation and performance; power system cost estimates; 10 MWe heat exchangers; 200 kWe heat exchanger articles; biofouling control;ammonia leak detection, and leak repair; rotating machinery; support subsystem; instrumentation and control; electrical subsystem; installation approach; net energy and resource analysis; and operability, maintainability, and safety are discussed. The conceptual design of the 40 MWe electrical power system includes four or five 10 MWe modules as designed for the 10 MWe pilot plant. (WHK)

  18. Methodology for the preliminary design of high performance schools in hot and humid climates

    Science.gov (United States)

    Im, Piljae

    A methodology to develop an easy-to-use toolkit for the preliminary design of high performance schools in hot and humid climates was presented. The toolkit proposed in this research will allow decision makers without simulation knowledge easily to evaluate accurately energy efficient measures for K-5 schools, which would contribute to the accelerated dissemination of energy efficient design. For the development of the toolkit, first, a survey was performed to identify high performance measures available today being implemented in new K-5 school buildings. Then an existing case-study school building in a hot and humid climate was selected and analyzed to understand the energy use pattern in a school building and to be used in developing a calibrated simulation. Based on the information from the previous step, an as-built and calibrated simulation was then developed. To accomplish this, five calibration steps were performed to match the simulation results with the measured energy use. The five steps include: (1) Using an actual 2006 weather file with measured solar radiation, (2) Modifying lighting & equipment schedule using ASHRAE's RP-1093 methods, (3) Using actual equipment performance curves (i.e., scroll chiller), (4) Using the Winkelmann's method for the underground floor heat transfer, and (5) Modifying the HVAC and room setpoint temperature based on the measured field data. Next, the calibrated simulation of the case-study K-5 school was compared to an ASHRAE Standard 90.1-1999 code-compliant school. In the next step, the energy savings potentials from the application of several high performance measures to an equivalent ASHRAE Standard 90.1-1999 code-compliant school. The high performance measures applied included the recommendations from the ASHRAE Advanced Energy Design Guides (AEDG) for K-12 and other high performance measures from the literature review as well as a daylighting strategy and solar PV and thermal systems. The results show that the net

  19. Calculation of aerodynamics of aerosol filter designs for cleaning of heavy liquid metal cooler reactor gas loops

    International Nuclear Information System (INIS)

    Valery P Melnikov; Pyotr N Martynov; Albert K Papovyants; Ivan V Yagodkin

    2005-01-01

    Full text of publication follows: One of the basic performances of aerosol filters is the aerodynamic resistance to the flow of gaseous medium to be cleaned. Calculation of the aerodynamics of aerosol filters in reference to the gas loops of reactor installations with heavy liquid metal coolant (HLMC) allows the design of the structural components of filters to be optimized to provide minimum initial resistance values. It is established that owing to various factors aerosol particles of different concentration and disperse composition are present always in the gas spaces of heavy liquid metal cooled reactor gas loops. To prevent the negative effect of aerosols on the equipment of the gas loops, it is reasonable to use filters of multistep design with sections of preliminary and fine cleaning to catch micron and submicron particles, respectively. A computer program and technique have been developed to evaluate the aerodynamics of folded aerosol filters for different parameters of their structural components, taking account of the aerosol spectrum and concentration. The algorithm of the calculation is presented by the example of a two-step design assembled in single vessel; the filter dimensions and pattern of the air flow to be cleaned are determined under the given boundary conditions. The evaluation of the aerodynamic resistance of filters was performed with consideration for local resistances and resistances of all the structural components of the filter (sudden constriction, expansion, the flow in air channels, filtering material and so on). Correlations have been derived for the resistance of air channels, filtering materials of preliminary and fine cleaning sections as a function of such parameters as the section depth (50-500 mm), the height of separators (3,5-20 mm), the filtering surface area (1,5-30 m 2 ). Based on the calculation results, the auto-similarity domain was brought out for the minimal values of filter resistances as a function of the ratio of

  20. Preliminary design study of pebble bed reactor HTR-PM base using once-through-then-out fuel recirculation

    International Nuclear Information System (INIS)

    Topan Setiadipura; Jupiter S Pane; Zuhair

    2016-01-01

    Pebble Bed Reactor (PBR) is one of the advanced reactor type implementing strong passive safety feature. In this type of design has the potential to do a cogeneration useful for the treatment of various minerals in various islands in Indonesia. The operation of the PBR can be simplified by implementing once-through-then-out (OTTO) fuel recirculation scheme in which pebble fuel only pass the core once time. The purpose of this research is to understand quantitative influence of the changing of fuel element recirculation on the PBR core performance and to find preliminary optimization design of PBR type reactor with OTTO recirculation scheme. PEBBED software was used to find PBR equilibrium core. The calculation result gives quantitative data on the impact of implementing a different fuel recirculation, especially using OTTO scheme. Furthermore, an early optimized PBR design based on HTR-PM using OTTO scheme was obtained where the power must be downgraded into 115 MWt in order to preserve the safety feature. The simplicity of the reactor operation and the reduction of reactor component with OTTO scheme still make this early optimized design an interesting alternative design, despite its power reduction from the reference design. (author)

  1. Sludge Treatment Project Engineered Container Retrieval And Transfer System Preliminary Design Hazard Analysis Supplement 1

    International Nuclear Information System (INIS)

    Franz, G.R.; Meichle, R.H.

    2011-01-01

    This 'What/If' Hazards Analysis addresses hazards affecting the Sludge Treatment Project Engineered Container Retrieval and Transfer System (ECRTS) NPH and external events at the preliminary design stage. In addition, the hazards of the operation sequence steps for the mechanical handling operations in preparation of Sludge Transport and Storage Container (STSC), disconnect STSC and prepare STSC and Sludge Transport System (STS) for shipping are addressed.

  2. Solid Waste Operations Complex W-113: Project cost estimate. Preliminary design report. Volume IV

    International Nuclear Information System (INIS)

    1995-01-01

    This document contains Volume IV of the Preliminary Design Report for the Solid Waste Operations Complex W-113 which is the Project Cost Estimate and construction schedule. The estimate was developed based upon Title 1 material take-offs, budgetary equipment quotes and Raytheon historical in-house data. The W-113 project cost estimate and project construction schedule were integrated together to provide a resource loaded project network

  3. Preliminary Design of Monitoring and Control Subsystem for GNSS Ground Station

    Directory of Open Access Journals (Sweden)

    Seongkyun Jeong

    2008-06-01

    Full Text Available GNSS (Global Navigation Satellite System Ground Station monitors navigation satellite signal, analyzes navigation result, and uploads correction information to satellite. GNSS Ground Station is considered as a main object for constructing GNSS infra-structure and applied in various fields. ETRI (Electronics and Telecommunications Research Institute is developing Monitoring and Control subsystem, which is subsystem of GNSS Ground Station. Monitoring and Control subsystem acquires GPS and Galileo satellite signal and provides signal monitoring data to GNSS control center. In this paper, the configurations of GNSS Ground Station and Monitoring and Control subsystem are introduced and the preliminary design of Monitoring and Control subsystem is performed. Monitoring and Control subsystem consists of data acquisition module, data formatting and archiving module, data error correction module, navigation solution determination module, independent quality monitoring module, and system operation and maintenance module. The design process uses UML (Unified Modeling Language method which is a standard for developing software and consists of use-case modeling, domain design, software structure design, and user interface structure design. The preliminary design of Monitoring and Control subsystem enhances operation capability of GNSS Ground Station and is used as basic material for detail design of Monitoring and Control subsystem.

  4. Preliminary design needs for pilot plant of Monazite processing into Thorium Oxide (ThO_2)

    International Nuclear Information System (INIS)

    Hafni Lissa Nuri; Prayitno; Abdul Jami; M-Pancoko

    2014-01-01

    Data and information collection aimed in order to meet the needs of the initial design for pilot plant of monazite processing into thorium oxide (ThO_2). The content of thorium in monazite is high in Indonesia between 2.9 to 4.1% and relatively abundant in Bangka Belitung Islands. Thorium can be used as fuel because of its potential is more abundant instead of uranium. Plant of thorium oxide commercially from monazite established starting from pilot uranium. Plant of thorium oxide commercially from monazite established starting from pilot plant in order to test laboratory data. Pilot plant design started from initial design, basic design, detailed design, procurement and construction. Preliminary design needs includes data feed and products, a block diagram of the process, a description of the process, the determination of process conditions and type of major appliance has been conducted. (author)

  5. General description of preliminary design of an experimental fusion reactor and the future problems

    International Nuclear Information System (INIS)

    Sako, Kiyoshi

    1976-01-01

    Recently, the studies on plasma physics has progressed rapidly, and promising experimental data emerged successively. Especially expectation mounts high that Tokamak will develop into power reactors. In Japan, the construction of large plasma devices such as JT-60 of JAERI is going to start, and after several years, the studies on plasma physics will come to the end of first stage, then the main research and development will be directed to power reactors. The studies on the design of practical fusion reactors have been in progress since 1973 in JAERI, and the preliminary design is being carried out. The purposes of the preliminary design are the clarification of the concept of the experimental reactor and the requirements for the studies on core plasma, the examination of the problems for developing main components and systems of the reactor, and the development of design technology. The experimental reactor is the quasi-steady reactor of 100 MW fusion reaction output, and the conditions set for the design and the basis of their setting are explained. The outline of the design, namely core plasma, blankets, superconductive magnets and the shielding with them, vacuum wall, neutral particle injection heating device, core fuel supply and exhaust system, and others, is described. In case of scale-up the reactor structural material which can withstand neutron damage must be developed. (Kako, I.)

  6. Designing learning apparatus to promote twelfth grade students’ understanding of digital technology concept: A preliminary studies

    Science.gov (United States)

    Marlius; Kaniawati, I.; Feranie, S.

    2018-05-01

    A preliminary learning design using relay to promote twelfth grade student’s understanding of logic gates concept is implemented to see how well it’s to adopted by six high school students, three male students and three female students of twelfth grade. This learning design is considered for next learning of digital technology concept i.e. data digital transmition and analog. This work is a preliminary study to design the learning for large class. So far just a few researches designing learning design related to digital technology with relay. It may due to this concept inserted in Indonesian twelfth grade curriculum recently. This analysis is focus on student difficulties trough video analysis to learn the concept. Based on our analysis, the recommended thing for redesigning learning is: students understand first about symbols and electrical circuits; the Student Worksheet is made in more detail on the assembly steps to the project board; mark with symbols at points in certain places in the circuit for easy assembly; assembly using relays by students is enough until is the NOT’s logic gates and the others that have been assembled so that effective time. The design of learning using relays can make the relay a liaison between the abstract on the digital with the real thing of it, especially in the circuit of symbols and real circuits. Besides it is expected to also enrich the ability of teachers in classroom learning about digital technology.

  7. Development of an Exploration-Class Cascade Distillation System: Flight Like Prototype Preliminary Design

    Science.gov (United States)

    Callahan, Michael R.; Sargusingh, Miriam J.

    2015-01-01

    The ability to recover and purify water through physiochemical processes is crucial for realizing long-term human space missions, including both planetary habitation and space travel. Because of their robust nature, distillation systems have been actively pursued as one of the technologies for water recovery. One such technology is the Cascade Distillation System (CDS) a multi-stage vacuum rotary distiller system designed to recover water in a microgravity environment. Its rotating cascading distiller operates similarly to the state of the art (SOA) vapor compressor distiller (VCD), but its control scheme and ancillary components are judged to be straightforward and simpler to implement into a successful design. Through the Advanced Exploration Systems (AES) Life Support Systems (LSS) Project, the NASA Johnson Space Center (JSC) in collaboration with Honeywell International is developing a second generation flight forward prototype (CDS 2.0). The key objectives for the CDS 2.0 design task is to provide a flight forward ground prototype that demonstrates improvements over the SOA system in the areas of increased reliability and robustness, and reduced mass, power and volume. It will also incorporate exploration-class automation. The products of this task are a preliminary flight system design and a high fidelity prototype of an exploration class CDS. These products will inform the design and development of the third generation CDS which is targeted for on-orbit DTO. This paper details the preliminary design of the CDS 2.0.

  8. European passive plant program preliminary safety analyses to support system design

    International Nuclear Information System (INIS)

    Saiu, Gianfranco; Barucca, Luciana; King, K.J.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

  9. Preliminary design package for residential heating/cooling system: Rankine air conditioner redesign

    Science.gov (United States)

    1978-01-01

    A summary of the preliminary redesign and development of a marketable single family heating and cooling system is presented. The interim design and schedule status of the residential (3-ton) redesign, problem areas and solutions, and the definition of plans for future design and development activities were discussed. The proposed system for a single-family residential heating and cooling system is a single-loop, solar-assisted, hydronic-to-warm-air heating subsystem with solar-assisted domestic water heating and a Rankine-driven expansion air-conditioning subsystem.

  10. Heat recovery and seed recovery development project: preliminary design report (PDR)

    Energy Technology Data Exchange (ETDEWEB)

    Arkett, A. H.; Alexander, K. C.; Bolek, A. D.; Blackman, B. K.; Kurrle, P. E.; Tram, S. V.; Warren, A. M.; Ziobrowski, A. J.

    1981-06-01

    The preliminary design and performance characteristics are described of the 20 MWt heat recovery and seed recovery (HRSR) system to be fabricated, installed, and evaluated to provide a technological basis for the design of commercial size HRSR systems for coal-fired open-cycle MHD power plants. The system description and heat and material balances, equipment description and functional requirements, controls, interfacing systems, and operation and maintenance are detailed. Appendices include: (1) recommended environmental requirements for compliance with federal and state of Tennessee regulations, (2) channel and diffuser simulator, (3) equipment arrangement drawings, and (4) channel and diffuser simulator barrel drawings. (WHK)

  11. Preliminary design package for residential heating/cooling system--Rankine air conditioner redesign

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    This report contains a summary of the preliminary redesign and development of a marketable single-family heating and cooling system. The objectives discussed are the interim design and schedule status of the Residential (3-ton) redesign, problem areas and solutions, and the definition of plans for future design and development activities. The proposed system for a single-family residential heating and cooling system is a single-loop, solar-assisted, hydronic-to-warm-air heating subsystem with solar-assisted domestic water heating and a Rankine-driven expansion air-conditioning subsystem.

  12. Power calculations using exact data simulation: A useful tool for genetic study designs

    NARCIS (Netherlands)

    van der Sluis, S.; Dolan, C.V.; Neale, M.C.; Posthuma, D.

    2008-01-01

    Statistical power calculations constitute an essential first step in the planning of scientific studies. If sufficient summary statistics are available, power calculations are in principle straightforward and computationally light. In designs, which comprise distinct groups (e.g., MZ & DZ twins),

  13. The MSFC Collaborative Engineering Process for Preliminary Design and Concept Definition Studies

    Science.gov (United States)

    Mulqueen, Jack; Jones, David; Hopkins, Randy

    2011-01-01

    This paper describes a collaborative engineering process developed by the Marshall Space Flight Center's Advanced Concepts Office for performing rapid preliminary design and mission concept definition studies for potential future NASA missions. The process has been developed and demonstrated for a broad range of mission studies including human space exploration missions, space transportation system studies and in-space science missions. The paper will describe the design team structure and specialized analytical tools that have been developed to enable a unique rapid design process. The collaborative engineering process consists of integrated analysis approach for mission definition, vehicle definition and system engineering. The relevance of the collaborative process elements to the standard NASA NPR 7120.1 system engineering process will be demonstrated. The study definition process flow for each study discipline will be will be outlined beginning with the study planning process, followed by definition of ground rules and assumptions, definition of study trades, mission analysis and subsystem analyses leading to a standardized set of mission concept study products. The flexibility of the collaborative engineering design process to accommodate a wide range of study objectives from technology definition and requirements definition to preliminary design studies will be addressed. The paper will also describe the applicability of the collaborative engineering process to include an integrated systems analysis approach for evaluating the functional requirements of evolving system technologies and capabilities needed to meet the needs of future NASA programs.

  14. Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Baik, Seung Joon; Bae, Seong Jun; Kim, Seong Gu; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO{sub 2} as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m{sup 2} and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase.

  15. Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

    International Nuclear Information System (INIS)

    Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

    2007-01-01

    In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

  16. Efficient preliminary floating offshore wind turbine design and testing methodologies and application to a concrete spar design.

    Science.gov (United States)

    Matha, Denis; Sandner, Frank; Molins, Climent; Campos, Alexis; Cheng, Po Wen

    2015-02-28

    The current key challenge in the floating offshore wind turbine industry and research is on designing economic floating systems that can compete with fixed-bottom offshore turbines in terms of levelized cost of energy. The preliminary platform design, as well as early experimental design assessments, are critical elements in the overall design process. In this contribution, a brief review of current floating offshore wind turbine platform pre-design and scaled testing methodologies is provided, with a focus on their ability to accommodate the coupled dynamic behaviour of floating offshore wind systems. The exemplary design and testing methodology for a monolithic concrete spar platform as performed within the European KIC AFOSP project is presented. Results from the experimental tests compared to numerical simulations are presented and analysed and show very good agreement for relevant basic dynamic platform properties. Extreme and fatigue loads and cost analysis of the AFOSP system confirm the viability of the presented design process. In summary, the exemplary application of the reduced design and testing methodology for AFOSP confirms that it represents a viable procedure during pre-design of floating offshore wind turbine platforms. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

  17. Preliminary design and thermal analysis of device for finish cooling Jaffa biscuits in a.d. 'Jaffa'- Crvenka

    Directory of Open Access Journals (Sweden)

    Salemović Duško R.

    2015-01-01

    Full Text Available In this paper preliminary design of device for finish cooling chocolate topping of biscuits in A.D. 'Jaffa'- Crvenka was done. The proposed preliminary design followed by the required technological process of finish cooling biscuits and required parameters of process which was supposed to get and which represented part of project task. Thermal analysis was made and obtained percentage error between surface contact of the air and chocolate topping, obtained from heat balance and geometrical over proposed preliminary design, wasn't more than 0.67%. This is a preliminary design completely justified because using required length of belt conveyor receive required temperature of chocolate topping at the end of the cooling process.

  18. Rapid Preliminary Design of Interplanetary Trajectories Using the Evolutionary Mission Trajectory Generator

    Science.gov (United States)

    Englander, Jacob

    2016-01-01

    Preliminary design of interplanetary missions is a highly complex process. The mission designer must choose discrete parameters such as the number of flybys, the bodies at which those flybys are performed, and in some cases the final destination. In addition, a time-history of control variables must be chosen that defines the trajectory. There are often many thousands, if not millions, of possible trajectories to be evaluated. This can be a very expensive process in terms of the number of human analyst hours required. An automated approach is therefore very desirable. This work presents such an approach by posing the mission design problem as a hybrid optimal control problem. The method is demonstrated on notional high-thrust chemical and low-thrust electric propulsion missions. In the low-thrust case, the hybrid optimal control problem is augmented to include systems design optimization.

  19. 1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1977-08-01

    This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations

  20. The Square Kilometre Array Science Data Processor. Preliminary compute platform design

    International Nuclear Information System (INIS)

    Broekema, P.C.; Nieuwpoort, R.V. van; Bal, H.E.

    2015-01-01

    The Square Kilometre Array is a next-generation radio-telescope, to be built in South Africa and Western Australia. It is currently in its detailed design phase, with procurement and construction scheduled to start in 2017. The SKA Science Data Processor is the high-performance computing element of the instrument, responsible for producing science-ready data. This is a major IT project, with the Science Data Processor expected to challenge the computing state-of-the art even in 2020. In this paper we introduce the preliminary Science Data Processor design and the principles that guide the design process, as well as the constraints to the design. We introduce a highly scalable and flexible system architecture capable of handling the SDP workload

  1. 1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    1977-08-01

    This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations. (DLC)

  2. Preliminary structural design of composite main rotor blades for minimum weight

    Science.gov (United States)

    Nixon, Mark W.

    1987-01-01

    A methodology is developed to perform minimum weight structural design for composite or metallic main rotor blades subject to aerodynamic performance, material strength, autorotation, and frequency constraints. The constraints and load cases are developed such that the final preliminary rotor design will satisfy U.S. Army military specifications, as well as take advantage of the versatility of composite materials. A minimum weight design is first developed subject to satisfying the aerodynamic performance, strength, and autorotation constraints for all static load cases. The minimum weight design is then dynamically tuned to avoid resonant frequencies occurring at the design rotor speed. With this methodology, three rotor blade designs were developed based on the geometry of the UH-60A Black Hawk titanium-spar rotor blade. The first design is of a single titanium-spar cross section, which is compared with the UH-60A Black Hawk rotor blade. The second and third designs use single and multiple graphite/epoxy-spar cross sections. These are compared with the titanium-spar design to demonstrate weight savings from use of this design methodology in conjunction with advanced composite materials.

  3. Design calculations for MAW storage experiment in the Asse salt mine

    International Nuclear Information System (INIS)

    Nipp, H.K.

    1987-01-01

    Several thermal pre-calculations examine what heat release is necessary in the storage experiment in order to produce a temperature level relevant to final storage at the edge of the borehole. It was found that the initial power must be 350-400 W/m. The thermo-mechanical design calculations are done on symmetrical equivalent models, in order to avoid a genuine spatial calculation. The results of the calculations show that the recovery of the radioactive waste is guaranteed over the whole experimental period, as the selected convergence space of 11 or 14 cm is sufficiently large. From the rock mechanics point of view, the MAW storage experiment is designed for the 800 m seam of the Asse mine, as no critical stresses are expected from calculations for the area of the borehole. (orig./RB) [de

  4. Censoring Data for Resistance Factor Calculations in Load and Resistance Factor Design: A Preliminary Study

    Science.gov (United States)

    James W. Evans; David W. Green

    2007-01-01

    Reliability estimates for the resistance distribution of wood product properties may be made from test data where all specimens are broken (full data sets) or by using data sets where information is obtained only from the weaker pieces in the distribution (censored data). Whereas considerable information exists on property estimation from full data sets, much less...

  5. Research Initiatives and Preliminary Results In Automation Design In Airspace Management in Free Flight

    Science.gov (United States)

    Corker, Kevin; Lebacqz, J. Victor (Technical Monitor)

    1997-01-01

    The NASA and the FAA have entered into a joint venture to explore, define, design and implement a new airspace management operating concept. The fundamental premise of that concept is that technologies and procedures need to be developed for flight deck and ground operations to improve the efficiency, the predictability, the flexibility and the safety of airspace management and operations. To that end NASA Ames has undertaken an initial development and exploration of "key concepts" in the free flight airspace management technology development. Human Factors issues in automation aiding design, coupled aiding systems between air and ground, communication protocols in distributed decision making, and analytic techniques for definition of concepts of airspace density and operator cognitive load have been undertaken. This paper reports the progress of these efforts, which are not intended to definitively solve the many evolving issues of design for future ATM systems, but to provide preliminary results to chart the parameters of performance and the topology of the analytic effort required. The preliminary research in provision of cockpit display of traffic information, dynamic density definition, distributed decision making, situation awareness models and human performance models is discussed as they focus on the theme of "design requirements".

  6. CAREM 25: actual status of the core neutronic design. Calculation line

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    This work follows the one titled 'Criteria for the CAREM 25 reactor core design. Neutronic aspects' presented at this congress, gives in detail the typical values regarding the core defined at this point. Besides, the neutronic calculation line used for the CAREM 25 reactor design is presented. (Author) [es

  7. Revised design calculations of lift systems; Elevator no setsubi keikaku ni okeru kotsu keisan

    Energy Technology Data Exchange (ETDEWEB)

    Maeda, T.; Komaya, K. [Mitsubishi Electric Corp., Tokyo (Japan)

    1998-07-01

    For sufficient transportation capacity and passengers comfort and convenience, it is very important to design the suitable lift systems (e.g., the appropriate number of cages, velocity, capacity etc.) using a model which describes real elevator movements. The procedure used in conventional design calculations for office buildings is to determine the transportation capacity for the up-peak traffic situation using a simple passengers arrival model. This paper presents a new design calculations using balanced traffic model, which can deal with the elevator movements considering passengers arrival rate. As some performance indexes to evaluate the quality of service can be calculated by using this model, lift system designers can determine the appropriate lift facilities as to satisfy their goals. The validity of the proposed model is also shown by comparing with the measured data in real lift systems. 6 refs., 9 figs., 2 tabs.

  8. Calculation of anti-seismic design for Xi'an pulsed reactor

    International Nuclear Information System (INIS)

    Li Shuian

    2002-01-01

    The author describes the reactor safety rule, safety regulation and design code that must be observed to anti-seismic design in Xi'an pulsed reactor. It includes the classification of reactor installation, determination of seismic loads, calculate contents, program, method, results and synthetically evaluation. According to the different anti-seismic structure character of reactor installation, an appropriate method was selected to calculate the seismic response. The results were evaluated synthetically using the design code and design requirement. The evaluate results showed that the anti-seismic design function of reactor installation of Xi'an pules reactor is well, and the structure integrality and normal property of reactor installation can be protect under the designed classification of the earthquake

  9. Preliminary safety assessment study for the conceptual design of a repository in tuff at Yucca Mountain

    International Nuclear Information System (INIS)

    Jackson, J.L.; Gram, H.F.; Hong, K.J.; Ng, H.S.; Pendergrass, A.M.

    1984-12-01

    Preliminary estimates of the upper bounds on postulated worst-case radiological releases resulting from possible accidents during the operating period of a prospective repository in tuff at Yucca Mountain are presented. Possible disrupting events are screened to identify the accidents of greatest potential consequence. The radiological dose commitments for the general public and repository personnel are estimated for postulated releases caused by natural phenomena, man-made events, and operational accidents. All postulated worst-case releases result in doses to the public that are lower than the 0.5-rem, whole-body dose-per-accident limit set by the Nuclear Regulatory Commission (NRC) in 10 CFR 60. Doses to repository personnel are within the NRC's 5.0-rem/yr occupational exposure limit set in 10 CFR 20 for normal operations. Doses are within this limit for all accidents except the transportation accident and fire in a drift. A preliminary risk assessment has also been performed. Based on this preliminary safety study, the proposed site boundaries and design criteria routinely used in constructing nuclear facilities appear to be adequate to protect the safety of the general public during the operating phase of the repository

  10. Towards a preliminary design of the ITER plasma control system architecture

    International Nuclear Information System (INIS)

    Treutterer, W.; Rapson, C.J.; Raupp, G.; Snipes, J.; Vries, P. de; Winter, A.; Humphreys, D.A.; Walker, M.; Tommasi, G. de; Cinque, M.; Bremond, S.; Moreau, P.; Nouailletas, R.; Felton, R.

    2017-01-01

    Highlights: • ITER control requirements and use scenarios for initial plasma operation have been analysed. • Basic choices from conceptual design could be confirmed. • Architectural design considers dynamic structure changes. • All PCS components are integrated in an exception handling hierarchy. - Abstract: Design of the ITER plasma control system is proceeding towards its next – preliminary design – stage. During the conceptual design in 2013 an overall assessment of high-level control tasks and their relationships has been conducted. The goal of the preliminary design is to show, that a reasonable implementation of the proposed concepts exists which fulfills the high-level requirements and is suitable for realistic use cases. This verification is conducted with focus on the concrete use cases of early operation and first plasma, since these phases are mandatory for ITER startup. In particular, detailed control requirements and functions for commissioning and first plasma operation including breakdown, burn-through and ramp-up in L-mode, as well as for planned or exceptional shutdown are identified. Control functions related to those operational phases and the underlying control system architecture are modeled. The goal is to check whether the flexibility of the conceptual architectural approach is adequate also in consideration of the more elaborate definitions for control functions and their interactions. In addition, architecture shall already be prepared for extension to H-mode operation and burn-control, even if the related control functions are only roughly defined at the moment. As a consequence, the architectural design is amended where necessary and converted into base components and infrastructure services allowing to deploy control and exception handling algorithms for the concrete first-plasma operation.

  11. Towards a preliminary design of the ITER plasma control system architecture

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Rapson, C.J.; Raupp, G. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Snipes, J.; Vries, P. de; Winter, A. [ITER Organization, Route de Vinon sur Verdon, 13067 St Paul Lez Durance (France); Humphreys, D.A.; Walker, M. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Tommasi, G. de; Cinque, M. [CREATE/Università di Napoli Federico II, Napoli (Italy); Bremond, S.; Moreau, P.; Nouailletas, R. [Association CEA pour la Fusion Contrôlée, CEA Cadarache, 13108 St Paul les Durance (France); Felton, R. [CCFE Fusion Association, Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire, OX14 3DB (United Kingdom)

    2017-02-15

    Highlights: • ITER control requirements and use scenarios for initial plasma operation have been analysed. • Basic choices from conceptual design could be confirmed. • Architectural design considers dynamic structure changes. • All PCS components are integrated in an exception handling hierarchy. - Abstract: Design of the ITER plasma control system is proceeding towards its next – preliminary design – stage. During the conceptual design in 2013 an overall assessment of high-level control tasks and their relationships has been conducted. The goal of the preliminary design is to show, that a reasonable implementation of the proposed concepts exists which fulfills the high-level requirements and is suitable for realistic use cases. This verification is conducted with focus on the concrete use cases of early operation and first plasma, since these phases are mandatory for ITER startup. In particular, detailed control requirements and functions for commissioning and first plasma operation including breakdown, burn-through and ramp-up in L-mode, as well as for planned or exceptional shutdown are identified. Control functions related to those operational phases and the underlying control system architecture are modeled. The goal is to check whether the flexibility of the conceptual architectural approach is adequate also in consideration of the more elaborate definitions for control functions and their interactions. In addition, architecture shall already be prepared for extension to H-mode operation and burn-control, even if the related control functions are only roughly defined at the moment. As a consequence, the architectural design is amended where necessary and converted into base components and infrastructure services allowing to deploy control and exception handling algorithms for the concrete first-plasma operation.

  12. Judicial problems in connection with preliminary decision and construction design approval in nuclear licensing procedures

    International Nuclear Information System (INIS)

    Schmieder, K.

    1977-01-01

    Standardization in nuclear engineering makes two demands on a legal instrument which is to make this standardization possible and which is to promote standardization in the nuclear licensing practice: On the basis of just one licence for a constructional part or a component, its applicability in any number of subsequent facility licensing procedures has to be warranted, and by virtue of its binding effect, standardization has to create a sufficiently big confidence protection with manufacturers, constructioneers and operators to offer sufficiently effective incentives for standardization. The nuclear preliminary decision pursuant to section 7 a of the Atomic Energy Act in the form of the component preliminary decision appears to be unsuitable as a legal instrument for standardization, as the preliminary decision refers exclusively to the construction of a concrete facility. For standardization in reactor engineering, the construction design approval appears to be basically the proper legal instrument on account of its legal structure as well as its economic effect. Its binding effect encouters a limitation with regard to third parties in so far that this limitation could question again the binding effect in a subsequent site-dependent nuclear licence procedure. The legal structure of the extent of the binding effect, which is decisive for the suitability of the construction design approval, lies with the legislator. The following questions have to be regulated: Ought the applicant to have a legal claim on the granting of a construction design approval, or ought it to be at the discretion of the authorities, and secondly, the extent of the binding effect in terms of time on the basis of the fixation of a time limit, or on the basis of the possibility of subsequent conditions to be imposed, or the revocation. (orig./HP) [de

  13. Interactive Block Games for Assessing Children's Cognitive Skills: Design and Preliminary Evaluation

    Directory of Open Access Journals (Sweden)

    Kiju Lee

    2018-05-01

    Full Text Available Background: This paper presents design and results from preliminary evaluation of Tangible Geometric Games (TAG-Games for cognitive assessment in young children. The TAG-Games technology employs a set of sensor-integrated cube blocks, called SIG-Blocks, and graphical user interfaces for test administration and real-time performance monitoring. TAG-Games were administered to children from 4 to 8 years of age for evaluating preliminary efficacy of this new technology-based approach.Methods: Five different sets of SIG-Blocks comprised of geometric shapes, segmented human faces, segmented animal faces, emoticons, and colors, were used for three types of TAG-Games, including Assembly, Shape Matching, and Sequence Memory. Computational task difficulty measures were defined for each game and used to generate items with varying difficulty. For preliminary evaluation, TAG-Games were tested on 40 children. To explore the clinical utility of the information assessed by TAG-Games, three subtests of the age-appropriate Wechsler tests (i.e., Block Design, Matrix Reasoning, and Picture Concept were also administered.Results: Internal consistency of TAG-Games was evaluated by the split-half reliability test. Weak to moderate correlations between Assembly and Block Design, Shape Matching and Matrix Reasoning, and Sequence Memory and Picture Concept were found. The computational measure of task complexity for each TAG-Game showed a significant correlation with participants' performance. In addition, age-correlations on TAG-Game scores were found, implying its potential use for assessing children's cognitive skills autonomously.

  14. Parallel calculation of sensitivity derivatives for aircraft design using automatic differentiation

    Energy Technology Data Exchange (ETDEWEB)

    Bischof, C.H.; Knauff, T.L. Jr. [Argonne National Lab., IL (United States); Green, L.L.; Haigler, K.J. [National Aeronautics and Space Administration, Hampton, VA (United States). Langley Research Center

    1994-01-01

    Realistic multidisciplinary design optimization (MDO) of advanced aircraft using state-of-the-art computers is an extremely challenging problem from both the physical modelling and computer science points of view. In order to produce an efficient aircraft design, many trade-offs must be made among the various physical design variables. Similarly, in order to produce an efficient design scheme, many trade-offs must be made among the various MDO implementation options. In this paper, we examine the effects of vectorization and coarse-grained parallelization on the SD calculation using a representative example taken from a transonic transport design problem.

  15. Preliminary CFD Analysis for HVAC System Design of a Containment Building

    Energy Technology Data Exchange (ETDEWEB)

    Son, Sung Man; Choi, Choengryul [ELSOLTEC, Yongin (Korea, Republic of); Choo, Jae Ho; Hong, Moonpyo; Kim, Hyungseok [KEPCO Engineering and Construction, Gimcheon (Korea, Republic of)

    2016-10-15

    HVAC (Heating, Ventilation, Air Conditioning) system has been mainly designed based on overall heat balance and averaging concepts, which is simple and useful for designing overall system. However, such a method has the disadvantage that cannot predict the local flow and temperature distributions in a containment building. In this study, a CFD (Computational Fluid Dynamics) preliminary analysis is carried out to obtain detailed flow and temperature distributions in a containment building and to ensure that such information can be obtained via CFD analysis. This approach can be useful for hydrogen analysis in an accident related to hydrogen released into a containment building. In this study, CFD preliminary analysis has been performed to obtain the detailed information of the reactor containment building by using the CFD analysis techniques and to ensure that such information can be obtained via CFD analysis. We confirmed that CFD analysis can offer enough detailed information about flow patterns and temperature field and that CFD technique is a useful tool for HVAC design of nuclear power plants.

  16. Mountaineer Commercial Scale Carbon Capture and Storage Project Topical Report: Preliminary Public Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Guy Cerimele

    2011-09-30

    This Preliminary Public Design Report consolidates for public use nonproprietary design information on the Mountaineer Commercial Scale Carbon Capture & Storage project. The report is based on the preliminary design information developed during the Phase I - Project Definition Phase, spanning the time period of February 1, 2010 through September 30, 2011. The report includes descriptions and/or discussions for: (1) DOE's Clean Coal Power Initiative, overall project & Phase I objectives, and the historical evolution of DOE and American Electric Power (AEP) sponsored projects leading to the current project; (2) Alstom's Chilled Ammonia Process (CAP) carbon capture retrofit technology and the carbon storage and monitoring system; (3) AEP's retrofit approach in terms of plant operational and integration philosophy; (4) The process island equipment and balance of plant systems for the CAP technology; (5) The carbon storage system, addressing injection wells, monitoring wells, system monitoring and controls logic philosophy; (6) Overall project estimate that includes the overnight cost estimate, cost escalation for future year expenditures, and major project risks that factored into the development of the risk based contingency; and (7) AEP's decision to suspend further work on the project at the end of Phase I, notwithstanding its assessment that the Alstom CAP technology is ready for commercial demonstration at the intended scale.

  17. The preliminary design of real-time neutron fissile material monitoring system

    International Nuclear Information System (INIS)

    Shi Jun; Ren Zhongguo; Zhang Ming; Zhao Zhiping; Chen Qi

    2013-01-01

    In this paper we present the preliminary design to carry out real-time neutron fissile material monitoring system, The system includes hardware and data acquisition software. For the hardware, it is employed with He3 proportional tubes as neutron detectors, polyethylene as moderator, and, to achieve the remote counting, RM4036 counting modules are connected to the remote computer through the 485 ports. The software with real-time data display and storage, alarm and other functions are developed using Visual Basic 6.0. (authors)

  18. Ex-vessel core catcher design requirements and preliminary concepts evaluation

    International Nuclear Information System (INIS)

    Friedland, A.J.; Tilbrook, R.W.

    1974-01-01

    As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

  19. Preliminary neutron design of the flux flatter for silicon doping at the RA10

    International Nuclear Information System (INIS)

    Cintas, A.; Bazzana, S.

    2012-01-01

    The neutron transmutation doping of silicon (NTD) is one of the facilities under development for the RA10 project. In order to obtain high quality semiconductor, commercial requirements of NTD include achieving high axial and radial uniformity in the silicon targets. Axial uniformity is achieved locating a neutron screen around the Si ingot, obtaining a flat axial distribution of the dopant concentration. We present the neutron design of this screen, also known as flux flattener. MCNP5 was used to model the screen design. We have reached a satisfactory preliminary screen design after numerous iterations. The fluctuation in the axial distribution of the reaction capture rate ( 30 Si(n,γ) 31 Si) is under ≠1,5%, which is the required level by the semiconductor industry to accept the final product (author)

  20. Preliminary Design of a Synchronized Narrow Bandwidth FEL for Taiwan Light Source

    CERN Document Server

    Keung Lau Wai; Ching Fan, Tai; Zone Hsiao Feng; Tung Hsu Kuo; Hwang, Ching Shiang; Cheng Kuo Chin; Huei Luo Guo; Jen Wang Duan; Ping Wang Jau; Huey Wang Min

    2004-01-01

    Design study of a narrow line-width, high power IR-FEL facility has been carried out at NSRRC. This machine is designed to synchronize with the U9 undulator radiation of Taiwan Light Source and therefore provide new opportunity for chemical dynamics and condensed matter research. It has been proposed to use a super-conducting linac to provide a 60 MeV high quality electron beam to drive a 2.5-10 microns FEL oscillator with U5 undulator. Operating this linac in energy recovery mode will also be considered as an option to improve overall system effeciency and reduce heat loss and radiation dosage at the beam dump. Performance requirements and outcomes from this preliminary design study will be reported.

  1. Preliminary design of an energy-conversion unit of radiation-voltaic battery

    International Nuclear Information System (INIS)

    Yang Yuqing; Wang Guanquan; Hu Rui; Gao Hui; Liu Yebing; Zhang Huaming; Luo Shunzhong

    2010-01-01

    Based on the principle of radiation-voltaic effect, a preliminary energy-conversion unit of radiation-voltaic battery was designed. Three energy-conversion units were manufactured and their electric I-V properties under irradiation of solid sources of 63 Ni and 3 H were measured. The I-V curves were analyzed and some ideas for improvement were presented. It was found that the designed energy-conversion unit deteriorated dramatically under irradiation of 241 Am source. The best U oc and I sc gained under irradiation of 2.96 x 10 8 Bq 63 Ni were 0.267 V and 28.4 nA, and were 0.260 V and 62.8 nA under irradiation of a 5.09 x 10 9 Bq 3 H source. Further efforts are being made to improve the design. (authors)

  2. Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

    International Nuclear Information System (INIS)

    Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won

    2008-12-01

    Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process

  3. Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process.

  4. V/STOL tilt rotor aircraft study. Volume 2: Preliminary design of research aircraft

    Science.gov (United States)

    1972-01-01

    A preliminary design study was conducted to establish a minimum sized, low cost V/STOL tilt-rotor research aircraft with the capability of performing proof-of-concept flight research investigations applicable to a wide range of useful military and commercial configurations. The analysis and design approach was based on state-of-the-art methods and maximum use of off-the-shelf hardware and systems to reduce development risk, procurement cost and schedules impact. The rotors to be used are of 26 foot diameter and are the same as currently under construction and test as part of NASA Tilt-Rotor Contract NAS2-6505. The aircraft has a design gross weight of 12,000 lbs. The proposed engines to be used are Lycoming T53-L-13B rated at 1550 shaft horsepower which are fully qualified. A flight test investigation is recommended which will determine the capabilities and limitations of the research aircraft.

  5. Preliminary physical design of 7 MeV proton RFQ for the accelerator driven-energy system

    International Nuclear Information System (INIS)

    Luo Zihua

    2000-01-01

    The preliminary physical design of 7 MeV proton RFQ for the ADS (Accelerator Driven-energy System) is briefly described. The design features and the basic parameters and the design version of the RFQ are discussed. The matches between IS and RFQ and between RFQ and CCDTL/DTL are also discussed. The ideas of research for the RFQ are presented

  6. a basic approach to designing e calculator and c oach to designing

    African Journals Online (AJOL)

    eobe

    machine may have three main functions ree main functions ... most control systems using digital logic w implemented using individual ... OACH TO DESIGNING EMBEDDED SYSTEMS USING A SI. OACH TO .... Jumper Wires. No Units. 1 yard.

  7. Development of mathematical pediatric phantoms for internal dose calculations: designs, limitations, and prospects

    International Nuclear Information System (INIS)

    Cristy, M.

    1980-01-01

    Mathematical phantoms of the human body at various ages are employed with Monte Carlo radiation transport codes for calculation of photon specific absorbed fractions. The author has developed a pediatric phantom series based on the design of the adult phantom, but with explicit equations for each organ so that organ sizes and marrow distributions could be assigned properly. Since the phantoms comprise simple geometric shapes, predictive dose capability is limited when geometry is critical to the calculation. Hence, there is a demand for better phantom design in situations where geometry is critical, such as for external irradiation or for internal emitters with low energy photons. Recent advances in computerized axial tomography (CAT) present the potential for derivation of anatomical information, which is so critical to development of phantoms, and ongoing developmental work on compuer architecture to handle large arrays for Monte Carlo calculations should make complex-geometry dose calculations economically feasible within this decade

  8. Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

    International Nuclear Information System (INIS)

    Setiadipura, T.; Zuhair; Irwanto, D.

    2015-01-01

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

  9. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  10. Calculator: A Hardware Design, Math and Software Programming Project Base Learning

    Directory of Open Access Journals (Sweden)

    F. Criado

    2015-03-01

    Full Text Available This paper presents the implementation by the students of a complex calculator in hardware. This project meets hardware design goals, and also highly motivates them to use competences learned in others subjects. The learning process, associated to System Design, is hard enough because the students have to deal with parallel execution, signal delay, synchronization … Then, to strengthen the knowledge of hardware design a methodology as project based learning (PBL is proposed. Moreover, it is also used to reinforce cross subjects like math and software programming. This methodology creates a course dynamics that is closer to a professional environment where they will work with software and mathematics to resolve the hardware design problems. The students design from zero the functionality of the calculator. They are who make the decisions about the math operations that it is able to resolve it, and also the operands format or how to introduce a complex equation into the calculator. This will increase the student intrinsic motivation. In addition, since the choices may have consequences on the reliability of the calculator, students are encouraged to program in software the decisions about how implement the selected mathematical algorithm. Although math and hardware design are two tough subjects for students, the perception that they get at the end of the course is quite positive.

  11. Iterative User Interface Design for Automated Sequential Organ Failure Assessment Score Calculator in Sepsis Detection.

    Science.gov (United States)

    Aakre, Christopher Ansel; Kitson, Jaben E; Li, Man; Herasevich, Vitaly

    2017-05-18

    The new sepsis definition has increased the need for frequent sequential organ failure assessment (SOFA) score recalculation and the clerical burden of information retrieval makes this score ideal for automated calculation. The aim of this study was to (1) estimate the clerical workload of manual SOFA score calculation through a time-motion analysis and (2) describe a user-centered design process for an electronic medical record (EMR) integrated, automated SOFA score calculator with subsequent usability evaluation study. First, we performed a time-motion analysis by recording time-to-task-completion for the manual calculation of 35 baseline and 35 current SOFA scores by 14 internal medicine residents over a 2-month period. Next, we used an agile development process to create a user interface for a previously developed automated SOFA score calculator. The final user interface usability was evaluated by clinician end users with the Computer Systems Usability Questionnaire. The overall mean (standard deviation, SD) time-to-complete manual SOFA score calculation time was 61.6 s (33). Among the 24% (12/50) usability survey respondents, our user-centered user interface design process resulted in >75% favorability of survey items in the domains of system usability, information quality, and interface quality. Early stakeholder engagement in our agile design process resulted in a user interface for an automated SOFA score calculator that reduced clinician workload and met clinicians' needs at the point of care. Emerging interoperable platforms may facilitate dissemination of similarly useful clinical score calculators and decision support algorithms as "apps." A user-centered design process and usability evaluation should be considered during creation of these tools. ©Christopher Ansel Aakre, Jaben E Kitson, Man Li, Vitaly Herasevich. Originally published in JMIR Human Factors (http://humanfactors.jmir.org), 18.05.2017.

  12. Shielding calculations for the design of neutron radiography facility around PARR

    International Nuclear Information System (INIS)

    Ashraf, M.M.; Khan, A.R.

    1989-06-01

    Shielding calculations for neutron radiography facility, proposed to be established around PARR have been carried out using two group diffusion theory and shielding formulae. Gamma radiation penetration calculations have been carried out using simple attenuation methods. The fabrication and installation of the neutron radiography facility would provide the basis for designing a better collimating system and would help establish under water radiography facility for the inspection of highly radioactive materials and components etc. (orig./A.B.)

  13. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    Fernandez, J. L.; Lopez, J.

    1986-01-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

  14. Verification of design calculations of a PGNAA setup using nuclear track ejectors

    Energy Technology Data Exchange (ETDEWEB)

    Naqvi, A.A. E-mail: aanaqvi@kfupm.edu.sa; Fazal-ur-Rehman,; Nagadi, .M.; Maslehuddin, M.; Khateeb-ur-Rehman; Kidwai, S

    2004-02-01

    A rectangular moderator assembly has been designed for the PGNAA setup at ing Fahd University of Petroleum and Minerals (KFUPM). The design calculations of the rectangular moderator, which were obtained through Monte Carlo simulation, have been verified experimentally through thermal neutron field measurement using CR-39 nuclear track detectors (NTDs). These measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The thermal neutron yield was measured inside the sample volume of the rectangular moderator by two NTDs fixed at back and front end of the sample cavity. The good agreement between he experimental results and the results of the calculations shows useful application of NTDs in verification of design calculations of a PGNAA setup.

  15. Design and preliminary testing of a Bottom-Mounted Second Shutdown Drive Mechanism for the KJRR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sanghaun; Lee, Jin Haeng; Yoo, Yeon-Sik, E-mail: yooys@kaeri.re.kr; Cho, Yeong-Garp; Lee, Hyokwang; Sun, Jongoh; Ryu, Jeong Soo

    2016-10-15

    Highlights: • The basic design principle, features and characteristics of the BMSSDM for KJRR are described. • The current development status based on practical fabrications, performance tests, and evaluations is described. • We have verified that all of the BMSSDM components satisfied their design requirements. • All of the performance requirements are satisfied from the performance test results. • The endurance test results show there are no structural failures and the wear of the impact parts in the hydraulic cylinder assembly is negligible. - Abstract: The KiJang Research Reactor (KJRR) is now being designed and undergoing preliminary construction by the Korea Atomic Energy Research Institute (KAERI). The driving parts of the Second Shutdown Drive Mechanism (SSDM) for the KJRR are located in a Reactivity Control Mechanism (RCM) room below the reactor pool bottom. In this paper, the design principle and concept of the Bottom-Mounted SSDM (BMSSDM) for the KJRR are introduced. From the experimental evaluations of the design, fabrication and performance, we verified that all of the BMSSDM components in the current design and development status satisfy their design requirements.

  16. Summary report for ITER Task -- D4: Activation calculations for the stainless steel ITER design

    International Nuclear Information System (INIS)

    Attaya, H.

    1995-02-01

    Detailed activation analysis for ITER has been performed as a part of ITER Task D4. The calculations have been performed for the shielding blanket (SS/water) and for the breeding blanket (LiN) options. The activation code RACC-P, which has been modified under IFER Task-D-10 for pulsed operation, has been used in this analysis. The spatial distributions of the radioactive inventory, decay heat, biological hazard potential, and the contact dose were calculated for the two designs for different operation modes and targeted fluences. A one-dimensional toroidal geometrical model has been utilized to determine the neutron fluxes in the two designs. The results are normalized for an inboard and outboard neutron wall loadings of 0.91 and 1.2 MW/M 2 , respectively. The point-wise distributions of the decay gamma sources have been calculated everywhere in the reactor at several times after the shutdown of the two designs and are then used in the transport code ONEDANT to calculate the biological dose everywhere in the reactor. The point-wise distributions of all the responses have also been calculated. These calculations have been performed for neutron fluences of 3.0 MWa/M 2 , which corresponds to the target fluence of ITER, and 0.1 MWa/M 2 , which is anticipated to correspond to the beginning of an extended maintenance period

  17. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  18. Calculation Of A Lattice Physics Parameter For SBWR Fuel Bundle Design

    International Nuclear Information System (INIS)

    Sardjono, Y.

    1996-01-01

    The maximum power peaking factor for Nuclear Power Plant SBWR type is 1.5. The precision for that calculation is related with the result of unit cell analysis each rod in the fuel bundles. This analysis consist of lattice eigenvalue, lattice average diffusion cross section as well as relative power peaking factor in the fuel rod for each fuel bundles. The calculation by using TGBLA computer code which is based on the transport and 168 group diffusion theory. From this calculation can be concluded that the maximum relative power peaking factor is 1.304 and lower than design limit

  19. To the problem of reinforced concrete reactor vessel design and calculation

    International Nuclear Information System (INIS)

    Kirillov, A.P.; Artem'ev, V.P.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Paushkin, A.G.

    1980-01-01

    Modern methods for calculating reactor vessels of prestressed reinforced concrete are analyzed. It is shown that during the stage of technical and economical substantiation of reactor vessel structure for determining its stressed-deformed state engineering methods of calculation must be used, in particular, fragmentation method, method of rings and plates, and during the stages of contract and detail designs - method of finite elements and dynamic relaxation method. It is concluded that when solving cyclic symmetrical problems as well as asymmetrical problems, calculational algorithms for axis-symmetrical distributions of stresses in the vessel with provision for elastic properties of structural material may be used

  20. Calculation and design of natural gas preheater equipments. Berechnung und Auslegung von Erdgas-Vorwaermeanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Fasold, H G [Ruhrgas AG, Essen (Germany); Wahle, H N [Ruhrgas AG, Essen (Germany)

    1994-04-01

    A greatly simplified model of a regulating station - consisting of the station components ''preheater'' and ''control unit'' - is used for the calculation and design of natural gas preheating plants. It is hereby possible to calculate the Joule-Thomson effect which occurs on the expansion of natural gas in the controller, the resulting drop in temperature and the thermal output required to compensate this which is to be supplied to the gas flow by the preheating plant. The calculation method and procedure are explained using a programming flowchart. The computational model presented was converted into a personal computer program, whose functioning is elucidated using a numerical example. (orig.)

  1. Preliminary design of the ITER AC/DC converters supplied by the Korean Domestic Agency

    International Nuclear Information System (INIS)

    Oh, J.S.; Choi, J.; Suh, J.H.; Liu, H.; Hwang, K.; Chung, I.; Lee, S.; Kang, J.; Park, H.; Jung, W.; Jo, S.; Gweon, H.; Lee, Y.; Lee, W.; Kim, J.B.; Han, S.H.; Hong, G.D.; Lee, J.S.; Lee, B.W.; Yeo, C.H.

    2013-01-01

    Highlights: ► A self-supporting aluminium structure and symmetrical thyristor assembly are devised to assure a strong and reliable ITER converter. ► Converters are designed to be installable in a compact space with three times higher power density than normal industrial installations. ► Heating of the building structure due to high magnetic field by converters are identified and certain solutions are addressed in the building design. ► A cooperative fast control scheme is adopted to compensate fast reactive power change of up to the level of 900 Mvar. -- Abstract: The preliminary design for ITER AC/DC converters under the responsibility of the Korean Domestic Agency is performed on the basis of the engineering experience of previous R and D for a full-scale 6-pulse CS (Central Solenoid) converter unit. This paper describes key features of the preliminary design for the respective sub-systems; integrated self-supporting aluminium structure and symmetrical thyristor assembly for strong and reliable converters, optimised impedance of the converter transformer to limit short circuit current, coaxial-type AC bus bars to shield high magnetic field around wall penetrations, compact components to fit into given building space. The insulation and the minimisation of electrical loops of concrete rebar below the converter installations are essential to prevent floor heating. Required output voltage or current of converters is provided by a conventional controller. A master controller is designed to collect predicted reactive powers from each converter and deliver processed data to the reactive power compensation (RPC) system to improve the regulation speed of the RPC controller with fast feed-forward compensation under fast reactive power transients

  2. A preliminary design of interior structure and foundation of an inflatable lunar habitat

    Science.gov (United States)

    Yin, Paul K.

    1989-01-01

    A preliminary structural design and analysis of an inflatable habitat for installation on the moon was completed. The concept takes the shape of a sphere with a diameter of approximately 16 meters. The interior framing provides five floor levels and is enclosed by a spherical air-tight membrane holding an interior pressure of 14.7 psi (101.4kpa). The spherical habitat is to be erected on the lunar surface with the lower one third below grade and the upper two thirds covered with a layer of lunar regolith for thermal insulation and shielding against radiation and meteoroids. The total dead weight (earth weight) of the structural aluminum, which is of vital interest for the costly space transportation, is presented. This structural dead weight represents a preliminary estimate without including structural details. The design results in two versions: one supports the weight of the radiation shielding in case of deflation of the fabric enclosure and the other assumes that the radiation shielding is self supporting. To gain some indication of the amount of structural materials needed if the identical habitat were installed on Mars and Earth, three additional design versions were generated where the only difference is in gravity. These additional design versions are highly academic since the difference will be much more than in gravity alone. The lateral loading due to dust storms on Mars and wind loads on Earth are some examples. The designs under the lunar gravity are realistic. They may not be adequate for final material procurement and fabrication, however, as the connection details, among other reasons, may effect the sizes of the structural members.

  3. Preliminary design of the ITER AC/DC converters supplied by the Korean Domestic Agency

    Energy Technology Data Exchange (ETDEWEB)

    Oh, J.S., E-mail: jsoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Choi, J.; Suh, J.H. [ITER Korea, National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Liu, H.; Hwang, K.; Chung, I.; Lee, S.; Kang, J.; Park, H.; Jung, W.; Jo, S.; Gweon, H.; Lee, Y.; Lee, W. [Dawonsys Corp., Siheung 429-450 (Korea, Republic of); Kim, J.B.; Han, S.H.; Hong, G.D.; Lee, J.S.; Lee, B.W.; Yeo, C.H. [Hyosung Corp., 450, Gongdeok-Dong, Seoul 121-720 (Korea, Republic of); and others

    2013-10-15

    Highlights: ► A self-supporting aluminium structure and symmetrical thyristor assembly are devised to assure a strong and reliable ITER converter. ► Converters are designed to be installable in a compact space with three times higher power density than normal industrial installations. ► Heating of the building structure due to high magnetic field by converters are identified and certain solutions are addressed in the building design. ► A cooperative fast control scheme is adopted to compensate fast reactive power change of up to the level of 900 Mvar. -- Abstract: The preliminary design for ITER AC/DC converters under the responsibility of the Korean Domestic Agency is performed on the basis of the engineering experience of previous R and D for a full-scale 6-pulse CS (Central Solenoid) converter unit. This paper describes key features of the preliminary design for the respective sub-systems; integrated self-supporting aluminium structure and symmetrical thyristor assembly for strong and reliable converters, optimised impedance of the converter transformer to limit short circuit current, coaxial-type AC bus bars to shield high magnetic field around wall penetrations, compact components to fit into given building space. The insulation and the minimisation of electrical loops of concrete rebar below the converter installations are essential to prevent floor heating. Required output voltage or current of converters is provided by a conventional controller. A master controller is designed to collect predicted reactive powers from each converter and deliver processed data to the reactive power compensation (RPC) system to improve the regulation speed of the RPC controller with fast feed-forward compensation under fast reactive power transients.

  4. Uranium Mill Tailings Remedial Action Project (UMTRAP), Slick Rock, Colorado, Revision 1. Volume 2, Calculations, Final design for construction

    International Nuclear Information System (INIS)

    1995-09-01

    Volume two contains calculations for: embankment design--slope stability analysis; embankment design--excavation stability; embankment design--settlement and cover cracking analysis; radon barrier design--statistical analysis of ra-226 concentrations for North Continent and Union Carbide sites; radon barrier design--RAECOM input data; radon barrier design--design thickness; and cover design--frost penetration depth

  5. Thermal Analysis of Iodine Satellite (iSAT) from Preliminary Design Review (PDR) to Critical Design Review (CDR)

    Science.gov (United States)

    Mauro, Stephanie

    2016-01-01

    The Iodine Satellite (iSAT) is a 12U cubesat with a primary mission to demonstrate the iodine fueled Hall Effect Thruster (HET) propulsion system. The spacecraft (SC) will operate throughout a one year mission in an effort to mature the propulsion system for use in future applications. The benefit of the HET is that it uses a propellant, iodine, which is easy to store and provides a high thrust-to-mass ratio. This paper will describe the thermal analysis and design of the SC between Preliminary Design Review (PDR) and Critical Design Review (CDR). The design of the satellite has undergone many changes due to a variety of challenges, both before PDR and during the time period discussed in this paper. Thermal challenges associated with the system include a high power density, small amounts of available radiative surface area, localized temperature requirements of the propulsion components, and unknown orbital parameters. The thermal control system is implemented to maintain component temperatures within their respective operational limits throughout the mission, while also maintaining propulsion components at the high temperatures needed to allow gaseous iodine propellant to flow. The design includes heaters, insulation, radiators, coatings, and thermal straps. Currently, the maximum temperatures for several components are near to their maximum operation limit, and the battery is close to its minimum operation limit. Mitigation strategies and planned work to solve these challenges will be discussed.

  6. Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

    2003-04-01

    In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed

  7. Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

    2003-04-01

    In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed.

  8. Preliminary power train design for a state-of-the-art electric vehicle

    Science.gov (United States)

    Ross, J. A.; Wooldridge, G. A.

    1978-01-01

    The state-of-the-art (SOTA) of electric vehicles built since 1965 was reviewed to establish a base for the preliminary design of a power train for a SOTA electric vehicle. The performance of existing electric vehicles were evaluated to establish preliminary specifications for a power train design using state-of-the-art technology and commercially available components. Power train components were evaluated and selected using a computer simulation of the SAE J227a Schedule D driving cycle. Predicted range was determined for a number of motor and controller combinations in conjunction with the mechanical elements of power trains and a battery pack of sixteen lead-acid batteries - 471.7 kg at 0.093 MJ/Kg (1040 lbs. at 11.7 Whr/lb). On the basis of maximum range and overall system efficiency using the Schedule D cycle, an induction motor and 3 phase inverter/controller was selected as the optimum combination when used with a two-speed transaxle and steel belted radial tires. The predicted Schedule D range is 90.4 km (56.2 mi). Four near term improvements to the SOTA were identified, evaluated, and predicted to increase range approximately 7%.

  9. Customer Relationship Management System in Occupational Safety & Health Companies: Research on Practice and Preliminary Design Solution

    Directory of Open Access Journals (Sweden)

    Robert Fabac

    2011-10-01

    Full Text Available One of the most prominent contemporary trends in formation of companies is the approach to development of a customer-oriented company. In this matter, various versions related to the intensity of this orientation are differentiated. Customer relationship management (CRM system is a well-known concept, and its practice is being studied and improved in connection to various sectors. Companies providing services of occupational safety and health (OHS mainly cooperate with a large number of customers and the quality of this cooperation largely affects the occupational safety and health of employees. Therefore, it is of both scientific and wider social interest to study and improve the relationship of these companies with their customers. This paper investigates the practice of applying CRM in Croatian OHS companies. It identifies the existing conditions and suggests possible improvements in the practice of CRM, based on experts’ assessments using analytic hierarchy process evaluation. Universal preliminary design was created as a framework concept for the formation of a typical customer-oriented OHS services company. Preliminary design includes a structural view, which provides more details through system diagrams, and an illustration of main cooperation processes of a company with its customer.

  10. Some preliminary design considerations for the ANS [Advanced Neutron Source] reactor cold source

    International Nuclear Information System (INIS)

    Henderson, D.L.

    1988-01-01

    Two areas concerned with the design of the Advanced Neutron Source (ANS) cold source have been investigated by simple one-dimensional calculations. The gain factors computed for a possible liquid nitrogen-15 cold source moderator are considerably below those computed for the much colder liquid deuterium moderator, as is reasonable considering the difference in moderator temperature. Nevertheless, nitrogen-15 does represent a viable option should safety related issues prohibit the use of deuterium as a moderating material. The slab geometry calculations have indicated that reflection of neutrons may be the dominant moderating mechanism and should be a consideration in the design of the cold source. 9 refs., 2 figs

  11. TPX: Contractor preliminary design review. Volume 1, Presentation and design description. Final report

    International Nuclear Information System (INIS)

    Hartman, D.; Naumovich; Walstrom, P.; Clarkson, I.; Schultheiss, J.; Burger, A.

    1995-01-01

    This first volume of the five volume set begins with a CPDR overview and then details the PF magnet system, manufacturing R ampersand D, Westinghouse R ampersand D, the central solenoid, the PF 5 ring coil, the PF 6/7 ring coil, quality assurance, and the system design description

  12. Preliminary design of a borax internal core-catcher for a gas cooled fast reactor

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Schumacher, G.

    1976-09-01

    Preliminary thermal calculations show that a core-catcher appears to be feasible, which is able to cope with the complete meltdown of the core and blankets of a 1,000 MWe GCFR. This core-catcher is based on borax (Na 2 B 4 O 7 ) as dissolving material of the oxide fuel and of the fission products occuring in oxide form. The borax is contained in steel boxes forming a 2.1 meter thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel, just underneath the reactor core. The fission products are dispersed in the pool formed by the liquid borax. The heat power density in the pool is conveniently reduced and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system. (orig.) [de

  13. Preliminary design study of an alternate heat source assembly for a Brayton isotope power system

    Science.gov (United States)

    Strumpf, H. J.

    1978-01-01

    Results are presented for a study of the preliminary design of an alternate heat source assembly (HSA) intended for use in the Brayton isotope power system (BIPS). The BIPS converts thermal energy emitted by a radioactive heat source into electrical energy by means of a closed Brayton cycle. A heat source heat exchanger configuration was selected and optimized. The design consists of a 10 turn helically wound Hastelloy X tube. Thermal analyses were performed for various operating conditions to ensure that post impact containment shell (PICS) temperatures remain within specified limits. These limits are essentially satisfied for all modes of operation except for the emergency cooling system for which the PICS temperatures are too high. Neon was found to be the best choice for a fill gas for auxiliary cooling system operation. Low cycle fatigue life, natural frequency, and dynamic loading requirements can be met with minor modifications to the existing HSA.

  14. Preliminary design of the beam screen cooling for the Future Circular Collider of hadron beams

    Science.gov (United States)

    Kotnig, C.; Tavian, L.

    2015-12-01

    Following recommendations of the recent update of the European strategy in particle physics, CERN has undertaken an international study of possible future circular colliders beyond the LHC. This study considers an option for a very high energy (100 TeV) hadron-hadron collider located in a quasi-circular underground tunnel having a circumference of 80 to 100 km. The synchrotron radiation emitted by the high-energy hadron beam increases by more than two orders of magnitude compared to the LHC. To reduce the entropic load on the superconducting magnets’ refrigeration system, beam screens are indispensable to extract the heat load at a higher temperature level. After illustrating the decisive constraints of the beam screen's refrigeration design, this paper presents a preliminary design of the length of a continuous cooling loop comparing helium and neon, for different cooling channel geometries with emphasis on the cooling length limitations and the exergetic efficiency.

  15. Preliminary design of the beam screen cooling for the Future Circular Collider of hadron beams

    CERN Document Server

    Kotnig, C

    2015-01-01

    Following recommendations of the recent update of the European strategy in particle physics, CERN has undertaken an international study of possible future circular colliders beyond the LHC. This study considers an option for a very high energy (100 TeV) hadron-hadron collider located in a quasi-circular underground tunnel having a circumference of 80 to 100 km. The synchrotron radiation emitted by the high-energy hadron beam increases by more than two orders of magnitude compared to the LHC. To reduce the entropic load on the superconducting magnets' refrigeration system, beam screens are indispensable to extract the heat load at a higher temperature level. After illustrating the decisive constraints of the beam screen's refrigeration design, this paper presents a preliminary design of the length of a continuous cooling loop comparing helium and neon, for different cooling channel geometries with emphasis on the cooling length limitations and the exergetic efficiency.

  16. TPX: Contractor preliminary design review. Volume 5, Manufacturing R ampersand D

    International Nuclear Information System (INIS)

    Roach, J.F.; Urban, W.M.; Hartman, D.

    1995-01-01

    TPX Insulation ampersand Impregnation R ampersand D test results are reported for 1x2 samples designed for screening candidate conduit insulation systems for TPX PF and TF coils. The epoxy/glass insulation system and three proposed alternate insulation systems employing Kapton, was evaluated in as received sample condition and after 10 thermal cycles in liquid nitrogen. Two DGBA impregnation systems, Shell 826 and CTD101K were investigated. Square incoloy 908 and 316 LN stainless hollow conduits were used for 1x2 sample fabrication. Capacitance, dielectric loss, and insulation resistance dielectric characteristics were measured for all samples. Partial discharge performance was measured for samples either in air, under silicon oil, or under liquid nitrogen up to 10kVrms at 60 Hz. Hipot screening was performed at 10 kVdc. The samples were cross sectioned and evaluated for impregnation quality. The implications of the test results on the TPX preliminary design decision are discussed

  17. Preliminary design studies of the draining tanks for the Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Allibert, M.; Heuer, D.; Brovchenko, M.; Laureau, A.; Ghetta, V.; Rubiolo, P.

    2014-01-01

    reactor called the Molten Salt Fast Reactor (MSFR). The reference MSFR design is a 3000 MWth reactor with a total fuel salt volume of 18 m3, operated at a mean fuel temperature of 750 deg. C. The first confinement barrier of the reactor includes a salt draining system. In case of a planned reactor shut down or in case of accidents leading to an excessive increase of the temperature in the fuel circuit, the fuel configuration may be changed passively by gravitational draining of the fuel salt in dedicated draining tank located under the reactor and designed to provide adequate reactivity margins while insuring a passive cooling of the fuel salt to extract the residual heat from the short to the long term. The present preliminary assessment of this sub-critical draining system has been performed to identify the physical constraints and to give some orders of magnitude of characteristic time periods (authors)

  18. TPX: Contractor preliminary design review. Volume 5, Manufacturing R&D

    Energy Technology Data Exchange (ETDEWEB)

    Roach, J.F.; Urban, W.M.; Hartman, D. [Everson Electric Co., Bekthlehem, PA (United States)

    1995-08-04

    TPX Insulation & Impregnation R&D test results are reported for 1x2 samples designed for screening candidate conduit insulation systems for TPX PF and TF coils. The epoxy/glass insulation system and three proposed alternate insulation systems employing Kapton, was evaluated in as received sample condition and after 10 thermal cycles in liquid nitrogen. Two DGBA impregnation systems, Shell 826 and CTD101K were investigated. Square incoloy 908 and 316 LN stainless hollow conduits were used for 1x2 sample fabrication. Capacitance, dielectric loss, and insulation resistance dielectric characteristics were measured for all samples. Partial discharge performance was measured for samples either in air, under silicon oil, or under liquid nitrogen up to 10kVrms at 60 Hz. Hipot screening was performed at 10 kVdc. The samples were cross sectioned and evaluated for impregnation quality. The implications of the test results on the TPX preliminary design decision are discussed.

  19. Ocean Thermal Energy Conservation (OTEC) power system development (PDS) II. Preliminary design report

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-10

    This report documents the results and conclusions of the PDS II, Phase I, preliminary design of a 10 MWe OTEC power system, using enhanced plate type heat exchangers, and of representative 0.2 MWe test articles. It further provides the documentation (specifications, drawings, trade studies, etc.) resulting from the design activities. The data and discussions of the technical concepts are organized to respond to the PDS II, Phase II proposal evaluation criteria. This volume, which specifically addresses the three evaluation categories (heat exchangers, rotating machinery, and power system configuration and performance) is an integral part of the Phase II plans (proposal) which describe the technical approach to delivering test articles to OTEC-1. In addition, there is a section which addresses power system cost and net energy analysis and another which discusses the results of stainless steel feasibility studies. Supporting documentation is contained in two appendix volumes.

  20. LASL experimental engineered waste burial facility: design considerations and preliminary plan

    International Nuclear Information System (INIS)

    DePoorter, G.L.

    1980-01-01

    The LASL Experimental Engineered Waste Burial Facility is a part of the National Low-Level Waste Management Program on Shallow-Land Burial Technology. It is a test facility where basic information can be obtained on the processes that occur in shallow-land burial operations and where new concepts for shallow-land burial can be tested on an accelerated basis on an appropriate scale. The purpose of this paper is to present some of the factors considered in the design of the facility and to present a preliminary description of the experiments that are initially planned. This will be done by discussing waste management philosophies, the purposes of the facility in the context of the waste management philosophy for the facility, and the design considerations, and by describing the experiments initially planned for inclusion in the facility, and the facility site

  1. Optimization study and preliminary design for Latina NPP early core retrieval and reactor dismantling

    International Nuclear Information System (INIS)

    Macci, E.; Zirpolo, S.; Imparato, A.; Cacace, A.; Parry, D.; Walkden, P.

    2002-01-01

    In June 2000, an agreement was established between Sogin and BNFL to enable the two companies to co-operate, using their specific experiences in the decommissioning field, for the benefit of projects in Italy, the United Kingdom and for third markets. A decommissioning strategy for the Latina NPP was initially developed in a Phase 1 Study which produced a conceptual design for the decommissioning of the reactor. This study was completed in June 2000. Since then, a second study has been completed, which has further developed the strategy and produced preliminary designs for the early dismantling of the core and reactor building at Latina. The engineering and safety data were produced in order to support Sogin in the preparation of a safety case for plant decommissioning. This safety case was submitted to the Italian Regulator, ANPA, in February 2002. (author)

  2. Calculation Sheet for the Basic Design of the ATLAS Fluid System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; Moon, S. K.; Yun, B. J.; Kwon, T. S.; Choi, K. Y.; Cho, S.; Park, C. K.; Lee, S. J.; Kim, Y. S.; Song, C. H.; Baek, W. P.; Hong, S. D

    2007-03-15

    The basic design of an integral effect test loop for pressurized water reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been carried out by Thermal-Hydraulics Safety Research Team in Korea Atomic Energy Research Institute (KAERI). The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400, and is scaled for full pressure and temperature conditions. This report includes calculation sheets for the basic design of ATLAS fluid systems, which are consisted of a reactor pressure vessel with core simulator, the primary loop piping, a pressurizer, reactor coolant pumps, steam generators, the secondary system, the safety system, the auxiliary system, and the heat loss compensation system. The present calculation sheets will be used to help understanding the basic design of the ATLAS fluid system and its based scaling methodology.

  3. Calculation Sheet for the Basic Design of the ATLAS Fluid System

    International Nuclear Information System (INIS)

    Park, Hyun Sik; Moon, S. K.; Yun, B. J.; Kwon, T. S.; Choi, K. Y.; Cho, S.; Park, C. K.; Lee, S. J.; Kim, Y. S.; Song, C. H.; Baek, W. P.; Hong, S. D.

    2007-03-01

    The basic design of an integral effect test loop for pressurized water reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been carried out by Thermal-Hydraulics Safety Research Team in Korea Atomic Energy Research Institute (KAERI). The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400, and is scaled for full pressure and temperature conditions. This report includes calculation sheets for the basic design of ATLAS fluid systems, which are consisted of a reactor pressure vessel with core simulator, the primary loop piping, a pressurizer, reactor coolant pumps, steam generators, the secondary system, the safety system, the auxiliary system, and the heat loss compensation system. The present calculation sheets will be used to help understanding the basic design of the ATLAS fluid system and its based scaling methodology

  4. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    International Nuclear Information System (INIS)

    Powers, J.

    2008-01-01

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials (1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF 4 or ThF 4 or some combination thereof. Future systems could look at using PuF 3 or PuF 4 as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory. Preliminary design studies

  5. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J

    2008-10-23

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials [1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF{sub 4} or ThF{sub 4} or some combination thereof. Future systems could look at using PuF{sub 3} or PuF{sub 4} as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory

  6. Tools for environmental simulations and calculations in an Integrated Design Process

    DEFF Research Database (Denmark)

    Petersen, Mads Dines; Knudstrup, Mary-Ann

    2010-01-01

    to address environmental issues. This paper in specific takes its starting point in a student project where the student is working with a building complex that have to fulfill the passive house standards and through that the student explores the use of the simulation and calculation tools in the design...... the possibilities of interoperability and the different possibilities are utilized pointing towards a heavier focus of the utilization of BIM tools in the design process from the first stages....

  7. Combining density functional theory calculations, supercomputing, and data-driven methods to design new materials (Conference Presentation)

    Science.gov (United States)

    Jain, Anubhav

    2017-04-01

    Density functional theory (DFT) simulations solve for the electronic structure of materials starting from the Schrödinger equation. Many case studies have now demonstrated that researchers can often use DFT to design new compounds in the computer (e.g., for batteries, catalysts, and hydrogen storage) before synthesis and characterization in the lab. In this talk, I will focus on how DFT calculations can be executed on large supercomputing resources in order to generate very large data sets on new materials for functional applications. First, I will briefly describe the Materials Project, an effort at LBNL that has virtually characterized over 60,000 materials using DFT and has shared the results with over 17,000 registered users. Next, I will talk about how such data can help discover new materials, describing how preliminary computational screening led to the identification and confirmation of a new family of bulk AMX2 thermoelectric compounds with measured zT reaching 0.8. I will outline future plans for how such data-driven methods can be used to better understand the factors that control thermoelectric behavior, e.g., for the rational design of electronic band structures, in ways that are different from conventional approaches.

  8. Impact of the measurement data on the CORD-2 nuclear design calculations of the NPP Krsko

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2004-01-01

    The CORD-2 package was developed at Jozef Stefan Institute and has been validated for the nuclear design calculations of PWR cores. It has been used for the independent verification of the NPP Krsko nuclear design for the last 6 cycles of operation. The accuracy of the package is very good fulfilling all criteria usually imposed on the design prediction of the reactor nuclear parameters. To obtain as robust package as possible and to eliminate potential systematic errors of the package, it was decided to rely on measured core power distributions. In core power measurements, which are performed each month of reactor operation, are used to obtain fuel assemblies burnup histories. Consequently, burnup distributions obtained from the power measurements of all previous cycles are taken as a starting point at the beginning of the considered cycle. Since a lot of experience has been gained with the package, it was decided to evaluate the impact of measurement data on the accuracy of the calculations. Burnup calculations of all 19 cycles of the NPP Krsko are repeated, building simultaneously the calculated library of burnup histories for all fuel assemblies. The basic reactor parameters such as HZP critical boron concentration, isothermal temperature coefficient, control rod worth and cycle length are compared to the results obtained with CORD-2 standard sequence of calculation and direct measurements.(author)

  9. Multigrid Algorithms for the Fast Calculation of Space-Charge Effects in Accelerator Design

    NARCIS (Netherlands)

    Pöplau, G.; Rienen, van U.; Geer, van der S.B.; Loos, de M.J.

    2004-01-01

    Numerical prediction of charged particle dynamics in accelerators is essential for the design and understanding of these machines. Methods to calculate the self-fields of the bunch, the so-called space-charge forces, become increasingly important as the demand for high-quality bunches increases. We

  10. Optimization Algorithms for Calculation of the Joint Design Point in Parallel Systems

    DEFF Research Database (Denmark)

    Enevoldsen, I.; Sørensen, John Dalsgaard

    1992-01-01

    In large structures it is often necessary to estimate the reliability of the system by use of parallel systems. Optimality criteria-based algorithms for calculation of the joint design point in a parallel system are described and efficient active set strategies are developed. Three possible...

  11. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-11-01

    The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

  12. Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

    2010-01-01

    To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

  13. Multilivel interfaces for power plant control rooms II: A preliminary design space

    International Nuclear Information System (INIS)

    Vicente, K.J.

    1992-01-01

    Events that are unfamiliar to operators and that have not been anticipated by designers pose the greatest threat to system safely in nuclear power plants. The abstraction hierarchy has been proposed as a representation frame-work that can be adopted to design interfaces that support operators in dealing with these unanticipated events. It consists of a multilevel representation format that represents a plant in terms of both physical and functional constraints. In a companion article, the work that has been done on this topic in academia, industry, and research laboratories was reviewed. On the basis of the results of that review, this article proposes a preliminary design space for multilevel interfaces based on the abstraction hierarchy. This space serves several worthwhile purposes: providing a unified framework within which to compare and contrast previous and future work in this area, providing a coherent research agenda by identifying some of the dimensions that can be meaningfully manipulated and evaluated in future experiments, and finally, serving as an input design by outlining the various decisions that need to be made in developing multilevel interfaces and the different options that are currently available for each of those decisions. Consequently this article should be of interest to researchers, designers, and regulators concerned with nuclear power-plant control rooms

  14. Manual for calculating critical loads of heavy metals for soils and surface waters; preliminary guidelines for environmental quality criteria, calculation methods and input data

    NARCIS (Netherlands)

    Vries, de W.; Bakker, D.J.

    1996-01-01

    Methodologies are described for calculating critical loads of lead, cadmium, copper, zinc, nickel, chromium and mercury for soils and surface waters. The aspects which are discussed are: selection of a computation model, determination of environmental-quality criteria for the metals, collection of

  15. Applications of thermodynamic calculations to Mg alloy design: Mg-Sn based alloy development

    International Nuclear Information System (INIS)

    Jung, In-Ho; Park, Woo-Jin; Ahn, Sang Ho; Kang, Dae Hoon; Kim, Nack J.

    2007-01-01

    Recently an Mg-Sn based alloy system has been investigated actively in order to develop new magnesium alloys which have a stable structure and good mechanical properties at high temperatures. Thermodynamic modeling of the Mg-Al-Mn-Sb-Si-Sn-Zn system was performed based on available thermodynamic, phase equilibria and phase diagram data. Using the optimized database, the phase relationships of the Mg-Sn-Al-Zn alloys with additions of Si and Sb were calculated and compared with their experimental microstructures. It is shown that the calculated results are in good agreement with experimental microstructures, which proves the applicability of thermodynamic calculations for new Mg alloy design. All calculations were performed using FactSage thermochemical software. (orig.)

  16. Jet-Surface Interaction: High Aspect Ratio Nozzle Test, Nozzle Design and Preliminary Data

    Science.gov (United States)

    Brown, Clifford; Dippold, Vance

    2015-01-01

    The Jet-Surface Interaction High Aspect Ratio (JSI-HAR) nozzle test is part of an ongoing effort to measure and predict the noise created when an aircraft engine exhausts close to an airframe surface. The JSI-HAR test is focused on parameters derived from the Turbo-electric Distributed Propulsion (TeDP) concept aircraft which include a high-aspect ratio mailslot exhaust nozzle, internal septa, and an aft deck. The size and mass flow rate limits of the test rig also limited the test nozzle to a 16:1 aspect ratio, half the approximately 32:1 on the TeDP concept. Also, unlike the aircraft, the test nozzle must transition from a single round duct on the High Flow Jet Exit Rig, located in the AeroAcoustic Propulsion Laboratory at the NASA Glenn Research Center, to the rectangular shape at the nozzle exit. A parametric nozzle design method was developed to design three low noise round-to-rectangular transitions, with 8:1, 12:1, and 16: aspect ratios, that minimizes flow separations and shocks while providing a flat flow profile at the nozzle exit. These designs validated using the WIND-US CFD code. A preliminary analysis of the test data shows that the actual flow profile is close to that predicted and that the noise results appear consistent with data from previous, smaller scale, tests. The JSI-HAR test is ongoing through October 2015. The results shown in the presentation are intended to provide an overview of the test and a first look at the preliminary results.

  17. Preliminary Design of Optimized Reactor Insulator for Severe Accident Mitigation of APR1400

    International Nuclear Information System (INIS)

    Heo, Sun; Lee, Jae-Gon; Kang, Yong-Chul

    2007-01-01

    APR1400, a Korean evolutionary advance light water reactor, has many advanced safety feature to prevent and mitigate of design basis accident (DBA) and severe accident. When reactor cooling system (RCS) fails to cooling its core, the core melted down and the molten core gathers together on bottom of reactor vessel. The molten core hurts reactor vessel and is released to containment, which raises the release of radioactive isotopes and the heating of the containment atmosphere. Finally, the corium is accumulated in the bottom of reactor cavity and it also raises the Molten Core and Concrete Interaction (MCCI) and the heating of containment atmosphere. There are two strategies to cooling molten core. Those are in-vessel retention and ex-vessel cooling. At the early stage of APR1400 design, only ex-vessel cooling which is cooling of the molten core outside the vessel after vessel failure is considered based on EPRI Utility Requirement Document (URD) for Evolutionary LWR. However, a need has been arisen to reflect current research findings on severe accident phenomena and mitigation technologies to Korean URD and IVRERVC (In-Vessel corium Retention using Ex-Reactor Vessel Cooling) was adopted APR1400. The ERVC is not considered as a licensing design basis but based on the defense-in-depth principle and safety margin basis, which is the top-tier requirement of the severe accident mitigation design as stated in the KURD. The Severe Accident Management strategy for APR1400 is intended to aid the plant operating staff to secure reactor vessel integrity in the early stage of the severe accident. As a part of a design implementation of IVR-ERVC for APR1400, we developed the preliminary design requirement, design specification and conceptual design

  18. MICE - Absorber and focus coil safety working group design document: Preliminary design and assessments

    International Nuclear Information System (INIS)

    Barr, Giles; Baynham, Elwyn; Black, Edgar; Bradshaw, Tom; Cummings, Mary Anne; Green, Michael A.; Ishimoto, Shigeru; Ivanyushenkov, Yury; Lau, Wing; Zisman, Michael

    2003-01-01

    A Neutrino Factory based on a muon storage ring is the ultimate tool for studies of neutrino oscillations, including possibly the discovery of leptonic CP violation. it is also the first step toward a muon collider. To develop a stored-muon-beam facility to serve as a Neutrino Factory, it is necessary to ''cool'' a muon beam (decrease its phase-space volume). The short lifetime of the muon, 2.2 (micro)s at rest, eliminates all currently demonstrated cooling techniques and requires that a new, heretofore untried, technique--ionization cooling--be employed. Although ionization cooling of muons has never been demonstrated in practice, it has been shown by end-to-end simulation and design studies to be an important factor both for the performance and for the cost of a Neutrino Factory. This motivates an international program of R and D, including an experimental demonstration at Rutherford Appleton Laboratory (RAL). The aims of the international Muon Ionization Cooling Experiment are: (1) to show that it is possible to design, engineer and build a section of cooling channel capable of giving the desired performance for a Neutrino Factory; and (2) to place it in a muon beam and measure its performance in various modes of operation and beam conditions, thereby investigating the limits and practicality of cooling. The MICE collaboration has designed an experiment in which a section of an ionization cooling channel is exposed to a muon beam. This cooling channel assembles liquid-hydrogen absorbers providing energy loss and high-gradient radio frequency (RF) cavities to re-accelerate the particles, all tightly contained in a magnetic channel. It reduces the beam transverse emittance by > 10% for muon momenta between 140 and 240 MeV/c. The layout of the experiment is shown. They utilize one complete magnetic cell of the cooling channel, comprising three absorber-focus-coil (AFC) modules and two RF-coupling-coil (RFCC) modules. Spectrometers placed before and after the

  19. Preliminary calculations for the CAFE project (Clean Air For Europe); Calculs preparatoires pour la strategie thematique CAFE (Clean Air For Europe)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-09-15

    The European Commission decided in 2001 an analysis program to reduce the atmospheric emissions. This report presents different limit scenari for France in 2020 (the reference scenari and the MTFR scenari, Maximum Technically Feasible Reduction), optimized scenari calculated by the RAINS model (Regional Air Pollution Information and Simulation), the costs of the scenari calculated with RAINS and the cost-benefit analysis of the strategy CAFE. From the study results, the benefits are higher than the costs, even with the most ambitious scenari. At an european level the emission reduction strategies have no effect on the employment but an impact on the Gross Domestic Product (decrease between 0,04 % and 0,12 % in function of the scenari). (A.L.B.)

  20. Preliminary Study of the Onset of Nucleate Boiling (ONB) for the Thermal-hydraulic Design of HANARO Irradiation non-instrumented Capsule during the Natural Convection

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Kyungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The HANARO reactor is an open-tank-in-pool type for easy access, and the capsules are being utilized for the irradiation test of materials and nuclear fuel in HANARO. The concept of the capsule is the direct contact with the coolant to cool the temperature of specimen down. To successfully accomplish the irradiation test, it is essential that the capsule should be designed considering the thermal margin such as the margin to Onset of Nucleate Boiling (ONB), the margin to Departure from Nucleate Boiling (DNB). In this paper, the preliminary study was performed by focusing on the ONB and the capsule design will be performed using the heat flux and temperature at ONB condition calculated in this paper. In this paper, the temperature and heat flux under ONB condition are simply calculated for the thermal design of fuel capsule for irradiation test. These values will be considered to design the non-instrumented capsule for natural circulation. To confirm the calculated value, detailed calculation will be performed using the one dimensional and multi-dimensional codes.

  1. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost

  2. The ICE spectrograph for PEPSI at the LBT: preliminary optical design

    Science.gov (United States)

    Pallavicini, Roberto; Zerbi, Filippo M.; Spano, Paolo; Conconi, Paolo; Mazzoleni, Ruben; Molinari, Emilio; Strassmeier, Klaus G.

    2003-03-01

    We present a preliminary design study for a high-resolution echelle spectrograph (ICE) to be used with the spectropolarimeter PEPSI under development at the LBT. In order to meet the scientific requirements and take full advantage of the peculiarities of the LBT (i.e. the binocular nature and the adaptive optics capabilities), we have designed a fiber-fed bench mounted instrument for both high resolution (R ≍ 100,000; non-AO polarimetric and integral light modes) and ultra-high resolution (R ≍ 300,000; AO integral light mode). In both cases, 4 spectra per order (two for each primary mirror) shall be accomodated in a 2-dimensional cross dispersed echelle format. In order to obtain a resolution-slit product of ≍ 100,000 as required by the science case, we have considered two alternative designs, one with two R4 echelles in series and the other with a sigle R4 echelle and fiber slicing. A white-pupil design, VPH cross-dispersers and two cameras of different focal length for the AO and non-AO modes are adopted in both cases. It is concluded that the single-echelle fiber-slicer solution has to be preferred in terms of performances, complexity and cost. It can be implemented at the LBT in two phases, with the long-camera AO mode added in a second phase depending on the availability of funds and the time-scale for implementation of the AO system.

  3. Drift design methodology and preliminary application for the Yucca Mountain Site Characterization Project

    International Nuclear Information System (INIS)

    Hardy, M.P.; Bauer, S.J.

    1991-12-01

    Excavation stability in an underground nuclear waste repository is required during construction, emplacement, retrieval (if required), and closure phases to ensure worker health and safety, and to prevent development of potential pathways for radionuclide migration in the post-closure period. Stable excavations are developed by appropriate excavation procedures, design of the room shape, design and installation of rock support reinforcement systems, and implementation of appropriate monitoring and maintenance programs. In addition to the loads imposed by the in situ stress field, the repository drifts will be impacted by thermal loads developed after waste emplacement and, periodically, by seismic loads from naturally occurring earthquakes and underground nuclear events. A priori evaluation of stability is required for design of the ground support system, to confirm that the thermal loads are reasonable, and to support the license application process. In this report, a design methodology for assessing drift stability is presented. This is based on site conditions, together with empirical and analytical methods. Analytical numerical methods are emphasized at this time because empirical data are unavailable for excavations in welded tuff either at elevated temperatures or under seismic loads. The analytical methodology incorporates analysis of rock masses that are systematically jointed, randomly jointed, and sparsely jointed. In situ thermal and seismic loads are considered. Methods of evaluating the analytical results and estimating ground support requirements for all the full range of expected ground conditions are outlines. The results of a preliminary application of the methodology using the limited available data are presented. 26 figs., 55 tabs

  4. Preliminary designs for 25 kWe advanced Stirling conversion systems for dish electric applications

    Science.gov (United States)

    Shaltens, Richard K.; Schreiber, Jeffrey G.

    Under the Department of Energy's (DOE) Solar Thermal Technology Program, Sandia National Laboratories is evaluating heat engines for terrestrial Solar Distributed Heat Receivers. The Stirling engine has been identified by Sandia as one of the most promising engines for terrestrial applications. The Stirling engine also has the potential to meet DOE's performance and cost goals. The NASA Lewis Research Center is conducting Stirling engine technology development activities directed toward a dynamic power source for space applications. Space power systems requirements include high reliability, very long life, low vibration and high efficiency. The free-piston Stirling engine has the potential for future high power space conversion systems, either nuclear or solar powered. Although both applications appear to be quite different, their requirements complement each other. Preliminary designs feature a free-piston Stirling engine, a liquid metal heat transport system, and a means to provide nominally 25 kW electric power to a utility grid while meeting DOE's performance and long term cost goals. The Cummins design incorporates a linear alternator to provide the electrical output, while the STC design generates electrical power indirectly through a hydraulic pump/motor coupled to an induction generator. Both designs for the ASCS's will use technology which can reasonably be expected to be available in the early 1990's.

  5. A preliminary study of mechanistic approach in pavement design to accommodate climate change effects

    Science.gov (United States)

    Harnaeni, S. R.; Pramesti, F. P.; Budiarto, A.; Setyawan, A.

    2018-03-01

    Road damage is caused by some factors, including climate changes, overload, and inappropriate procedure for material and development process. Meanwhile, climate change is a phenomenon which cannot be avoided. The effects observed include air temperature rise, sea level rise, rainfall changes, and the intensity of extreme weather phenomena. Previous studies had shown the impacts of climate changes on road damage. Therefore, several measures to anticipate the damage should be considered during the planning and construction in order to reduce the cost of road maintenance. There are three approaches generally applied in the design of flexible pavement thickness, namely mechanistic approach, mechanistic-empirical (ME) approach and empirical approach. The advantages of applying mechanistic approach or mechanistic-empirical (ME) approaches are its efficiency and reliability in the design of flexible pavement thickness as well as its capacity to accommodate climate changes in compared to empirical approach. However, generally, the design of flexible pavement thickness in Indonesia still applies empirical approach. This preliminary study aimed to emphasize the importance of the shifting towards a mechanistic approach in the design of flexible pavement thickness.

  6. Waste Receiving and Processing Facility Module 1: Volume 1, Preliminary Design report

    International Nuclear Information System (INIS)

    1992-03-01

    The Preliminary Design Report (Title 1) for the Waste Receiving and Processing (WRAP) Module 1 provides a comprehensive narrative description of the proposed facility and process systems, the basis for each of the systems design, and the engineering assessments that were performed to support the technical basis of the Title 1 design. The primary mission of the WRAP 1 Facility is to characterize and certify contact-handled (CH) waste in 55-gallon drums for disposal. Its secondary function is to certify CH waste in Standard Waste Boxes (SWBs) for disposal. The preferred plan consist of retrieving the waste and repackaging as necessary in the Waste Receiving and Processing (WRAP) facility to certify TRU waste for shipment to the Waste Isolation Pilot Plant (WIPP) in New Mexico. WIPP is a research and development facility designed to demonstrate the safe and environmentally acceptable disposal of TRU waste from National Defense programs. Retrieved waste found to be Low-Level Waste (LLW) after examination in the WRAP facility will be disposed of on the Hanford site in the low-level waste burial ground. The Hanford Site TRU waste will be shipped to the WIPP for disposal between 1999 and 2013

  7. Preliminary concept design of the divertor remote handling system for DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2014-11-15

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

  8. Preliminary Study on Kano Model in the Conceptual Design Activities for Product Lifecycle Improvement

    Science.gov (United States)

    Fahrul Hassan, Mohd; Rahman, M. R. A.; Arifin, A. M. T.; Ismail, A. E.; Rasidi Ibrahim, M.; Zulafif Rahim, M.; Fauzi Ahmad, Md

    2017-08-01

    Product manufactured with short life cycle had only one major issue, it can lead to increasing volume of waste. Day by day, this untreated waste had consumed many landfill spaces, waiting for any possible alternatives. Lack of product recovery knowledge and recyclability features imprinted into product design are one of the main reason behind all this. Sustainable awareness aspect should not just be implied into people’s mind, but also onto product design. This paper presents a preliminary study on Kano model method in the conceptual design activities to improve product lifecycle. Kano model is a survey-type method, used to analyze and distinguished product qualities or features, also how the customers may have perceived them. Three important attributes of Kano model are performance, attractive and must-be. The proposed approach enables better understanding of customer requirements while providing a way for Kano model to be integrated into engineering design to improve product’s end-of-life. Further works will be continued to provide a better lifecycle option (increase percentage of reuse, remanufacture or recycle, whereby decrease percentage of waste) of a product using Kano model approach.

  9. Calculational benchmark comparisons for a low sodium void worth actinide burner core design

    International Nuclear Information System (INIS)

    Hill, R.N.; Kawashima, M.; Arie, K.; Suzuki, M.

    1992-01-01

    Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational database, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified benchmark model of a metal fueled, low void worth actinide burner design is detailed; and two independent neutronic performance evaluations are compared. Calculated performance characteristics are evaluated for three spatially uniform compositions (fresh uranium/plutonium, batch-averaged uranium/transuranic, and batch-averaged uranium/transuranic with fission products) and a regional depleted distribution obtained from a benchmark depletion calculation. For each core composition, the flooded and voided multiplication factor, power peaking factor, sodium void worth (and its components), flooded Doppler coefficient and control rod worth predictions are compared. In addition, the burnup swing, average discharge burnup, peak linear power, and fresh fuel enrichment are calculated for the depletion case. In general, remarkably good agreement is observed between the evaluations. The most significant difference is predicted performance characteristics is a 0.3--0.5% Δk/(kk) bias in the sodium void worth. Significant differences in the transmutation rate of higher actinides are also observed; however, these differences do not cause discrepancies in the performing predictions

  10. Medical chilling device designed for hypothermic hydration graft storage system: Design, thermohydrodynamic modeling, and preliminary testing

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jung Hwan [Hongik University, Seoul (Korea, Republic of)

    2015-02-15

    Hypothermic hydration graft storage is essential to reduce the metabolic demand of cells in vitro. The alleviated metabolic demands reduce the emergence rate of anaerobic metabolism generating adenosine triphosphate (ATP) energy that creates free radicals. The cessive free radicals can damage cells and tissues due to their highly oxidative power with molecules. Current cooling systems such as a conventional air cooling system and an ice pack system are inappropriate for chilling cell tissues in vitro because of inconvenience in use and inconsistent temperature sustainability caused by large size and progressive melting, respectively. Here, we develop a medical chilling device (MCD) for hypothermic hydration graft storage based on thermo-hydrodynamic modeling and thermal electric cooling technology. Our analysis of obtained hydrodynamic thermal behavior of the MCD revealed that the hypothermic condition of 4 .deg. C was continuously maintained, which increased the survival rates of cells in vitro test by reduced free radicals. The validated performance of the MCD promises future development of an optimal hypothermic hydration graft storage system designed for clinical use.

  11. Preliminary neutronics design studies for a 400 MWt STAR-LM

    International Nuclear Information System (INIS)

    Aliberti, G.; Yang, W. S.; Stillman, J. A.; Hill, R. N.

    2004-01-01

    Neutronics design studies for a 400 MWt high temperature fast reactor are being performed, utilizing lead coolant, transuranic (TRU) nitride fuel, and HT-9 structural material. Under the main design constraints of long fuel lifetime, natural convection heat transport, semi-autonomous control, and small unit size, parametric studies were performed to maximize the discharge burnup and minimize the burnup reactivity swing. Based on the results of these parametric studies, two point designs were developed for a single-batch once-through fuel cycle; one is a 15 full power year cycle design with core volume of 9.5 cubic meters, and the other is a 12 full power year cycle design with core volume of 7.4 cubic meters. For these two point designs, fuel cycle analyses and reactivity feedback coefficients calculations were performed. The 9.5 cubic meter design achieved an average discharge burnup of 83 MWd/kg with a maximum reactivity change over the lifetime of 0.6%. The peak fast fluence was well within the fast fluence limit of HT9, and both average and peak power densities were well below the estimated limit for natural circulation. The performances of the 7.4 cubic meter design were slightly inferior to this design. To enhance the passive safety characteristics, however, further design improvements need to be made to reduce the coolant density coefficient and to increase the radial expansion coefficient. (authors)

  12. PROLIB: code to create production library of nuclear data for design calculations

    International Nuclear Information System (INIS)

    Wittkopf, W.A.; Tilford, J.M.; Furtney, M.

    1977-02-01

    The PROLIB program creates, updates, and edits the production library used in the B and W nuclear design system. The production library contains the material cross section data required to perform the thermal and epithermal spectrum calculations in the NULIF program. PROLIB collapses cross section data from the master libraries, produced by the ETOGM and THOR programs, to the desired production library group structures. The physics models that are used, the calculations that are performed in PROLIB, the input, and the output are described. Information that is required to use PROLIB along with a sample problem that illustrates the input and output formats and that provides a benchmark problem are given

  13. Shielding calculation techniques used in the design of fuel storage systems

    International Nuclear Information System (INIS)

    Wang, S.S.; Massey, J.V.

    1986-01-01

    This paper addresses the shielding design and analysis of a concrete modular spent fuel storage system. Particular attention is given to comparing various computation techniques in determining bulk shielding thickness, and also in dealing with the radiation streaming effect through the air exit penetration openings in the module. Three computer codes QADMOD, ANISN, and DOT-IV were used to solve the same problem. In addition, hand albedo calculation were done to augment the result of the QADMOD calculation to properly deal with the surface scattering

  14. Core physics design calculation of mini-type fast reactor based on Monte Carlo method

    International Nuclear Information System (INIS)

    He Keyu; Han Weishi

    2007-01-01

    An accurate physics calculation model has been set up for the mini-type sodium-cooled fast reactor (MFR) based on MCNP-4C code, then a detailed calculation of its critical physics characteristics, neutron flux distribution, power distribution and reactivity control has been carried out. The results indicate that the basic physics characteristics of MFR can satisfy the requirement and objectives of the core design. The power density and neutron flux distribution are symmetrical and reasonable. The control system is able to make a reliable reactivity balance efficiently and meets the request for long-playing operation. (authors)

  15. Prediction of the Main Engine Power of a New Container Ship at the Preliminary Design Stage

    Science.gov (United States)

    Cepowski, Tomasz

    2017-06-01

    The paper presents mathematical relationships that allow us to forecast the estimated main engine power of new container ships, based on data concerning vessels built in 2005-2015. The presented approximations allow us to estimate the engine power based on the length between perpendiculars and the number of containers the ship will carry. The approximations were developed using simple linear regression and multivariate linear regression analysis. The presented relations have practical application for estimation of container ship engine power needed in preliminary parametric design of the ship. It follows from the above that the use of multiple linear regression to predict the main engine power of a container ship brings more accurate solutions than simple linear regression.

  16. Hybrid Spectral Micro-CT: System Design, Implementation, and Preliminary Results

    CERN Document Server

    Bennett, James R; Xu, Qiong; Yu, Hengyong; Walsh, Michael; Butler, Anthony; Butler, Phillip; Cao, Guohua; Mohs, Aaron; Wang, Ge

    2014-01-01

    Spectral CT has proven an important development in biomedical imaging, and there have been several publications in the past years demonstrating its merits in pre-clinical and clinical applications. In 2012, Xu et al. reported that near-term implementation of spectral micro-CT could be enhanced by a hybrid architecture: a narrow-beam spectral "interior" imaging chain integrated with a traditional wide-beam "global" imaging chain. This hybrid integration coupled with compressive sensing (CS)-based interior tomography demonstrated promising results for improved contrast resolution, and decreased system cost and radiation dose. The motivation for the current study is implementation and evaluation of the hybrid architecture with a first-of-its-kind hybrid spectral micro-CT system. Preliminary results confirm improvements in both contrast and spatial resolution. This technology is shown to merit further investigation and potential application in future spectral CT scanner design.

  17. Preliminary design of betavoltaic battery using Co-60 and Pm-147 with GaAs substrate

    Energy Technology Data Exchange (ETDEWEB)

    Waris, A., E-mail: awaris@fi.itb.ac.id; Basar, K. [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Kusumawati, Y.; Alfarobi, A. S.; Aji, I. K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

    2016-03-11

    Battery is very important for the present daily life, especially for portable devices. The longer utilization time the better performance of battery. Betavoltaic battery is a device that converts energy from beta decays of radioactive nuclide into electric current. One of merits of the later battery is the life time that can be more than ten years without recharging. To develop the betavoltaic battery for energy source of portable devices we have performed a preliminary simulation design of betavoltaic battery using Pm-147 and Co-60 a beta emitter radionuclides with n-GaAs substrate. From the results we found that the combination of Pm-147 with n-GaAs substrate results in 9.0% of efficiency and higher output current compared to references.

  18. Experiments, conceptual design, preliminary cost estimates and schedules for an underground research facility

    International Nuclear Information System (INIS)

    Korbin, G.; Wollenberg, H.; Wilson, C.; Strisower, B.; Chan, T.; Wedge, D.

    1981-09-01

    Plans for an underground research facility are presented, incorporating techniques to assess the hydrological and thermomechanical response of a rock mass to the introduction and long-term isolation of radioactive waste, and to assess the effects of excavation on the hydrologic integrity of a repository and its subsequent backfill, plugging, and sealing. The project is designed to utilize existing mine or civil works for access to experimental areas and is estimated to last 8 years at a total cost for contruction and operation of $39.0 million (1981 dollars). Performing the same experiments in an existing underground research facility would reduce the duration to 7-1/2 years and cost $27.7 million as a lower-bound estimate. These preliminary plans and estimates should be revised after specific sites are identified which would accommodate the facility

  19. Preliminary design of betavoltaic battery using Co-60 and Pm-147 with GaAs substrate

    International Nuclear Information System (INIS)

    Waris, A.; Basar, K.; Kusumawati, Y.; Alfarobi, A. S.; Aji, I. K.

    2016-01-01

    Battery is very important for the present daily life, especially for portable devices. The longer utilization time the better performance of battery. Betavoltaic battery is a device that converts energy from beta decays of radioactive nuclide into electric current. One of merits of the later battery is the life time that can be more than ten years without recharging. To develop the betavoltaic battery for energy source of portable devices we have performed a preliminary simulation design of betavoltaic battery using Pm-147 and Co-60 a beta emitter radionuclides with n-GaAs substrate. From the results we found that the combination of Pm-147 with n-GaAs substrate results in 9.0% of efficiency and higher output current compared to references.

  20. The Science benefits and preliminary design of the southern hemisphere gravitational wave detector AIGO

    Energy Technology Data Exchange (ETDEWEB)

    Blair, D G; Barriga, P; Coward, D; Dumas, J-C; Fan, Y; Gras, S; Howell, E; Ju, L; Miao, H [School of Physics, University of Western Australia, Perth, WA 6009 (Australia); Brooks, A F; Hosken, D J; Munch, J; Veitch, P J [Department of Physics, The University of Adelaide, Adelaide, SA, 5005 Australia (Australia); Charlton, P [School of Computing and Mathematics, Charles Sturt University, NSW 2678 (Australia); Galloway, D [School of Mathematical Sciences, Monash University, Vic 3800 (Australia); Hughes, S [Department of Physics, Massachusetts Institute of Technology, Cambridge, MA 02139-4307 (United States); McClelland, D E; Scott, S M; Slagmolen, B J J [Department of Physics, Australian National University, Canberra, ACT 0200 (Australia); Melatos, A [School of Physics University of Melbourne, Parkville Vic 3010 Australia (Australia)], E-mail: dgb@physics.uwa.edu.au (and others)

    2008-07-15

    The proposed southern hemisphere gravitational wave detector AIGO increases the projected average baseline of the global array of ground based gravitational wave detectors by a factor {approx}4. This allows the world array to be substantially improved. The orientation of AIGO allows much better resolution of both wave polarisations. This enables better distance estimates for inspiral events, allowing unambiguous optical identification of host galaxies for about 25% of neutron star binary inspiral events. This can allow Hubble Law estimation without optical identification of an outburst, and can also allow deep exposure imaging with electromagnetic telescopes to search for weak afterglows. This allows independent estimates of cosmological acceleration and dark energy as well as improved understanding of the physics of neutron star and black hole coalescences. This paper reviews and summarises the science benefits of AIGO and presents a preliminary conceptual design.

  1. The Science benefits and preliminary design of the southern hemisphere gravitational wave detector AIGO

    International Nuclear Information System (INIS)

    Blair, D G; Barriga, P; Coward, D; Dumas, J-C; Fan, Y; Gras, S; Howell, E; Ju, L; Miao, H; Brooks, A F; Hosken, D J; Munch, J; Veitch, P J; Charlton, P; Galloway, D; Hughes, S; McClelland, D E; Scott, S M; Slagmolen, B J J; Melatos, A

    2008-01-01

    The proposed southern hemisphere gravitational wave detector AIGO increases the projected average baseline of the global array of ground based gravitational wave detectors by a factor ∼4. This allows the world array to be substantially improved. The orientation of AIGO allows much better resolution of both wave polarisations. This enables better distance estimates for inspiral events, allowing unambiguous optical identification of host galaxies for about 25% of neutron star binary inspiral events. This can allow Hubble Law estimation without optical identification of an outburst, and can also allow deep exposure imaging with electromagnetic telescopes to search for weak afterglows. This allows independent estimates of cosmological acceleration and dark energy as well as improved understanding of the physics of neutron star and black hole coalescences. This paper reviews and summarises the science benefits of AIGO and presents a preliminary conceptual design

  2. Design principle of TVO's final repository and preliminary adaptation to site specific conditions

    International Nuclear Information System (INIS)

    Salo, J-P.; Reikkola, R.

    1995-01-01

    Teollisuuden Voima Oy (TVO) is responsible for the management of spent fuel produced by the Olkiluoto power plant. TVO's current programme of spent fuel management is based on the guidelines and time schedule set by the Finnish Government. TVO has studied a final disposal concept in which the spent fuel bundles are encapsulated in copper canisters and emplaced in Finnish bedrock. According to the plan the final repository for spent fuel will be in operation by 2020. TVO's updated technical plans for the disposal of spent fuel together with a performance analysis (TVO-92) were submitted to the authorities in 1992. The paper describes the design principle of TVO's final repository and preliminary adaptation of the repository to site specific conditions. (author). 10 refs., 5 figs

  3. Microwave frequency sweep interferometer for plasma density measurements in ECR ion sources: Design and preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Torrisi, Giuseppe [INFN - Laboratori Nazionali del Sud, Via S. Sofia 62, 95125 Catania (Italy); University Mediterranea of Reggio Calabria, Reggio Calabria (Italy); Mascali, David; Neri, Lorenzo; Leonardi, Ornella; Celona, Luigi; Castro, Giuseppe; Agnello, Riccardo; Caruso, Antonio; Passarello, Santi; Longhitano, Alberto; Gammino, Santo [INFN - Laboratori Nazionali del Sud, Via S. Sofia 62, 95125 Catania (Italy); Sorbello, Gino [INFN - Laboratori Nazionali del Sud, Via S. Sofia 62, 95125 Catania (Italy); University of Catania, Catania, Italy and INFN-LNS, Catania (Italy); Isernia, Tommaso [University Mediterranea of Reggio Calabria, Reggio Calabria (Italy)

    2016-02-15

    The Electron Cyclotron Resonance Ion Sources (ECRISs) development is strictly related to the availability of new diagnostic tools, as the existing ones are not adequate to such compact machines and to their plasma characteristics. Microwave interferometry is a non-invasive method for plasma diagnostics and represents the best candidate for plasma density measurement in hostile environment. Interferometry in ECRISs is a challenging task mainly due to their compact size. The typical density of ECR plasmas is in the range 10{sup 11}–10{sup 13} cm{sup −3} and it needs a probing beam wavelength of the order of few centimetres, comparable to the chamber radius. The paper describes the design of a microwave interferometer developed at the LNS-INFN laboratories based on the so-called “frequency sweep” method to filter out the multipath contribution in the detected signals. The measurement technique and the preliminary results (calibration) obtained during the experimental tests will be presented.

  4. Design a computational program to calculate the composition variations of nuclear materials in the reactor operations

    International Nuclear Information System (INIS)

    Mohmmadnia, Meysam; Pazirandeh, Ali; Sedighi, Mostafa; Bahabadi, Mohammad Hassan Jalili; Tayefi, Shima

    2013-01-01

    Highlights: ► The atomic densities of light and heavy materials are calculated. ► The solution is obtained using Runge–Kutta–Fehlberg method. ► The material depletion is calculated for constant flux and constant power condition. - Abstract: The present work investigates an appropriate way to calculate the variations of nuclides composition in the reactor core during operations. Specific Software has been designed for this purpose using C#. The mathematical approach is based on the solution of Bateman differential equations using a Runge–Kutta–Fehlberg method. Material depletion at constant flux and constant power can be calculated with this software. The inputs include reactor power, time step, initial and final times, order of Taylor Series to calculate time dependent flux, time unit, core material composition at initial condition (consists of light and heavy radioactive materials), acceptable error criterion, decay constants library, cross sections database and calculation type (constant flux or constant power). The atomic density of light and heavy fission products during reactor operation is obtained with high accuracy as the program outputs. The results from this method compared with analytical solution show good agreements

  5. Preliminary Axial Flow Turbine Design and Off-Design Performance Analysis Methods for Rotary Wing Aircraft Engines. Part 2; Applications

    Science.gov (United States)

    Chen, Shu-cheng, S.

    2009-01-01

    In this paper, preliminary studies on two turbine engine applications relevant to the tilt-rotor rotary wing aircraft are performed. The first case-study is the application of variable pitch turbine for the turbine performance improvement when operating at a substantially lower shaft speed. The calculations are made on the 75 percent speed and the 50 percent speed of operations. Our results indicate that with the use of the variable pitch turbines, a nominal (3 percent (probable) to 5 percent (hypothetical)) efficiency improvement at the 75 percent speed, and a notable (6 percent (probable) to 12 percent (hypothetical)) efficiency improvement at the 50 percent speed, without sacrificing the turbine power productions, are achievable if the technical difficulty of turning the turbine vanes and blades can be circumvented. The second casestudy is the contingency turbine power generation for the tilt-rotor aircraft in the One Engine Inoperative (OEI) scenario. For this study, calculations are performed on two promising methods: throttle push and steam injection. By isolating the power turbine and limiting its air mass flow rate to be no more than the air flow intake of the take-off operation, while increasing the turbine inlet total temperature (simulating the throttle push) or increasing the air-steam mixture flow rate (simulating the steam injection condition), our results show that an amount of 30 to 45 percent extra power, to the nominal take-off power, can be generated by either of the two methods. The methods of approach, the results, and discussions of these studies are presented in this paper.

  6. Validation of the CORD-2 System for the Nuclear Design Calculations of the NPP Krsko Core

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2016-01-01

    The CORD-2 package intended for core design calculations of PWRs has be recently updated with some improved models. Since the modifications could substantially influence the obtained results, a technical validation process is required. This paper presents comparison of some calculated and measured parameters of the NPP Krsko core needed to qualify the package. Critical boron concentrations at hot full power for selected cycle burnup points and several parameters obtained during the start-up testing at the beginning of each cycle (hot zero power critical concentration, isothermal temperature coefficient and rods worth) for all 27 finished cycles of operation are considered. In addition, assembly-wise power distribution for some selected cycles is checked. Comparison has shown very good agreement of the CORD-2 calculated values with the selected measured parameter of the NPP Krsko core.(author).

  7. Evaluation of the performance of mini-WIMS in design calculations for SGHWR's

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1980-07-01

    In order to use the WIMS code for SGHWR design calculations it is desirable to reduce the computing time to a minimum. To this end, a study has been made of the effects of using condensed data libraries with few groups in the main transport routine and with coarse mesh representations. The results of initial lattice calculations are given in considerable detail for a set of SGHW experimental cores. The effects of condensation on attainable burnup and irradiated fuel composition for natural and enriched power reactor lattices have also been studied. Comparisons between detailed and condensed WIMS calculations are the main theme of the report but METHUSELAH and experimental results are included whenever possible. (author)

  8. Nuclear performance calculations for the ELMO Bumpy Torus Reactor (EBTR) reference design

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.

    1977-12-01

    The nuclear performance of the ELMO Bumpy Torus Reactor reference design has been calculated using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV transport cross-section data and nuclear response functions. The calculated results include estimates of the spatial and integral heating rate with emphasis on the recovery of fusion neutron energy in the blanket assembly and minimization of the energy deposition rates in the cryogenic magnet coil assemblies. The tritium breeding ratio in the natural lithium-laden blanket was calculated to be 1.29 tritium nuclei per incident neutron. The radiation damage in the reactor structural material and in the magnet assembly is also given

  9. Design and preliminary evaluation of an exoskeleton for upper limb resistance training

    Science.gov (United States)

    Wu, Tzong-Ming; Chen, Dar-Zen

    2012-06-01

    Resistance training is a popular form of exercise recommended by national health organizations, such as the American College of Sports Medicine (ACSM) and the American Heart Association (AHA). This form of training is available for most populations. A compact design of upper limb exoskeleton mechanism for homebased resistance training using a spring-loaded upper limb exoskeleton with a three degree-of-freedom shoulder joint and a one degree-of-freedom elbow joint allows a patient or a healthy individual to move the upper limb with multiple joints in different planes. It can continuously increase the resistance by adjusting the spring length to train additional muscle groups and reduce the number of potential injuries to upper limb joints caused by the mass moment of inertia of the training equipment. The aim of this research is to perform a preliminary evaluation of the designed function by adopting an appropriate motion analysis system and experimental design to verify our prototype of the exoskeleton and determine the optimal configuration of the spring-loaded upper limb exoskeleton.

  10. Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Solack; Carol Mason

    2012-03-01

    A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

  11. Preliminary worst-case accident analysis to support the conceptual design of a potential repository in tuff

    International Nuclear Information System (INIS)

    Jackson, J.L.; Gram, H.F.; Hong, K.J.; Ng, H.S.; Pendergrass, A.M.

    1984-01-01

    The Nevada Waste Storage Investigations (NNWSI) Project is conducting investigations to determine suitability of a site at Yucca Mountain for development as a high-level waste repository. In support of conceptual design, a preliminary analysis has been performed to identify events that could cause radiological releases from the surface facilities during the operations period. Accidental releases were modeled short-duration release plumes, dispersed under averaged climatic conditions, using the AIRDOS-EPA code. consequences of these accidents, in 50-yr integrated dose commitments to operations personnel, to the minimally exposed member of the public, and to the general population in the surrounding area were calculated. risk to the general public from each event was also assessed. All postulated accidents result in doses to pers of the public that are lower than the 0.5 rem/accident limit set by the NRC in 10 CFR 60. For those accidents that do not involve both fire and breach of waste canisters, doses to operations personnel are behind the NRC limit for routine operations of 5 rem/yr set in 10 CFR 20. Accidents that involve fire and breach of waste canisters may cause doses to some operations personnel that are in excess of this limit

  12. Preliminary worst-case accident analysis to support the conceptual design of a potential repository in tuff

    International Nuclear Information System (INIS)

    Jackson, J.L.; Gram, H.F.; Hong, K.J.; Ng, H.S.; Pendergrass, A.M.

    1984-01-01

    The Nevada Nuclear Waste Storage Investigations (NNWSI) Project is conducting investigations to determine the suitability of a site at Yucca Mountain for development as a high level waste repository. In support of the conceptual design, a preliminary analysis has been performed to identify events that could cause radiological releases from the surface facilities during the operations period. Accidental releases were modeled as short-duration release plumes, dispersed under averaged climatic conditions, using the AIRDOS-EPA code. The consequences of these accidents, in 50-yr integrated dose commitments to operations personnel, to the maximally exposed member of the public, and to the general population in the surrounding area were calculated. The risk to the general public from each event was also assessed. All postulated accidents result in doses to members of the public that are lower than the 0.5 rem/accident limit set by the NRC in 10 CFR 60. For those accidents that do not involve both fire and breach of waste canisters, doses to operations personnel are within the NRC limit for routine operations of 5 rem/yr set in 10 CFR 20. Accidents that involve fire and breach of waste canisters may cause doses to some operations personnel that are in excess of this limit. 18 references, 1 figure, 3 tables

  13. About the application of MCNP4 code in nuclear reactor core design calculations

    International Nuclear Information System (INIS)

    Svarny, J.

    2000-01-01

    This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

  14. Preliminary conceptual design and cost estimation for Korea Advanced Pyroprocessing Facility Plus (KAPF+)

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il, E-mail: nwiko@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Ho Hee, E-mail: nhhlee@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Choi, Sungyeol, E-mail: csy@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Sung-Ki, E-mail: sgkim1@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Byung Heung, E-mail: b.h.park@ut.ac.kr [Department of Chemical and Biological Engineering, Korea National University of Transportation, 50 Daehak-ro, Chungju-si, Chungbuk, 380-702 (Korea, Republic of); Lee, Hyo Jik, E-mail: hyojik@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, In Tae, E-mail: nitkim@kaeri.re.kr [Department of Chemical and Biological Engineering, Korea National University of Transportation, 50 Daehak-ro, Chungju-si, Chungbuk, 380-702 (Korea, Republic of); Lee, Han Soo, E-mail: hslee5@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2014-10-01

    Highlights: • Conceptual design is created for a pilot pyroprocessing plant treating PWR spent fuel. • Pilot-scale design is based on a capacity of 400 tHM/yr with 60 years lifetime. • All individual processes are integrated into a single system from feed to products. • Overall facility design is developed for a pilot pyroprocessing plant. • Unit process cost is estimated for pyroprocessing with uncertainties. - Abstract: Korea has developed pyroprocessing technology as a potential option for recycling spent fuels (SFs) from pressurized water reactors (PWRs). The pyroprocessing consists of various key unit processes and a number of research activities have been focused on each process. However, to realize the whole pyroprocessing concept, there is a critical need for integrating the individual developments and addressing a material flow from feed to final products. In addition, the advancement on overall facility design is an indispensable aspect for demonstration and commercialization of the pyroprocessing. In this study, a facility named as Korea Advanced Pyroprocess Facility Plus (KAPF+) is conceptualized with a capacity of 400 tHM/yr. The process steps are categorized based on their own characteristics while the capacities of process equipment are determined based on the current technical levels. The facility concept with a site layout of 104,000 m{sup 2} is developed by analyzing the operation conditions and materials treated in each process. As an economic approach to the proposed facility, the unit cost (781 $/kgHM denominated in 2009 USD) for KAPF+ is also analyzed with the conceptual design with preliminary sensitivity assessments including decontamination and decommissioning costs, a discount rate, staffing costs, and plant lifetime. While classifying and describing cost details of KAPF+, this study compares the unit cost of KAPF+ treating PWR SF to that of the pyroprocessing facility treating sodium-cooled fast reactor (SFR) SF.

  15. New shipyard layout design for the preliminary phase & case study for the green field project

    Science.gov (United States)

    Song, Young Joo; Woo, Jong Hun

    2013-03-01

    For several decades, Asian nations such as Korea, Japan and China have been leading the shipbuilding industry since the decline in Europe and America. However, several developing countries such as India, Brazil, etc. are going to make an entrance into the shipbuilding industry. These developing countries are finding technical partners or information providers because they are in situation of little experiences and technologies. Now, the shipbuilding engineering companies of shipbuilding advanced countries are getting a chance of engineering business against those developing countries. The starting point of this business model is green field project for the construction of new shipyard. This business model is started with a design of the shipyard layout. For the conducting of the shipyard layout design, four kinds of engineering parts are required. Those are civil engineering, building engineering, utility engineering and production layout engineering. Among these parts, production layout engineering is most important because its result is the foundation of the other engineering parts and it determines the shipyard capacity during the shipyard operation lifecycle. Previous researches about the shipyard layout design are out of the range from the business requirements because most research cases are in the tower of ivory, which means that there are little consideration of real ship and shipbuilding operation. In this paper, a shipyard layout design for preliminary phase is conducted for the target of newly planned shipyard at Venezuela of South America with an integrated method that is capable of dealing with actual master data from the shipyard. The layout design method of this paper is differentiated from the previous researches in that the actual product data from the target ship and the actual shipbuilding operation data are used for the required area estimation.

  16. New shipyard layout design for the preliminary phase & case study for the green field project

    Directory of Open Access Journals (Sweden)

    Young Joo Song

    2013-03-01

    Full Text Available For several decades, Asian nations such as Korea, Japan and China have been leading the shipbuilding industry since the decline in Europe and America. However, several developing countries such as India, Brazil, etc. are going to make an entrance into the shipbuilding industry. These developing countries are finding technical partners or information providers because they are in situation of little experiences and technologies. Now, the shipbuilding engineering companies of shipbuilding advanced countries are getting a chance of engineering business against those developing countries. The starting point of this business model is green field project for the construction of new shipyard. This business model is started with a design of the shipyard layout. For the conducting of the shipyard layout design, four kinds of engineering parts are required. Those are civil engineering, building engineering, utility engineering and production layout engineering. Among these parts, production layout engineering is most important because its result is the foundation of the other engineering parts and it determines the shipyard capacity during the shipyard operation lifecycle. Previous researches about the shipyard layout design are out of the range from the business requirements because most research cases are in the tower of ivory, which means that there are little consideration of real ship and shipbuilding operation. In this paper, a shipyard layout design for preliminary phase is conducted for the target of newly planned shipyard at Venezuela of South America with an integrated method that is capable of dealing with actual master data from the shipyard. The layout design method of this paper is differentiated from the previous researches in that the actual product data from the target ship and the actual shipbuilding operation data are used for the required area estimation.

  17. Influential parameters for designing and power consumption calculating of cumin mower

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoodi, E.; Jafari, A. [Tehran Univ., Karaj (Iran, Islamic Republic of). Dept. of Agricultural Machinery Engineering

    2010-07-01

    This paper reported on a study in which the consuming power and design of cumin mowers was calculated. The parameters required for calculating power consumption and designing of cumin mowers were measured along with some engineering properties of cumin stems. These included shearing and bending tests on cumin stem and specifying the coefficient of friction between mower knives and cumin stem. The relationships between static and dynamic friction forces being exerted on mower runners by soil with normal load were determined along with the factor affecting soil moisture. Some of the other parameters that are important for calculating the power consumption and design of an optimized mower include harvest moisture content; maximum and average of cumin stem diameter; maximum bio-yield point of force and maximum ultimate point of force in the cutting; average energy required to cut a stem; maximum elasticity module; maximum bending rupture force; average energy required for bending a stem; friction coefficient between the stem and knife edge; relation between bio-yield force, failure force, elasticity and diameter in the cutting; relation between rupture forces and diameter in the bending; and mower weight.

  18. Ocean thermal energy conversion power system development-I. Preliminary design report. Volume 3. Appendixes D, E, and F. Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-18

    The conceptual design of a 40 to 50 MW closed cycle ammonia OTEC commercial plant, the preliminary design of a 10 MW OTEC module analogous to the 50 MW module, and the preliminary design of heat exchanger test articles (evaporator and condenser) representative of the 50 MW heat exchangers for testing in OTEC-1 are presented. This volume includes the appendices: D) system equipment (hardware breakdown structure; 10-MW hardware listing; list of support and maintenance equipment, tools and spare parts; sacrificial anodes; M.A.N. brush; and Alclad 3004 data); E) heat exchanger supporting data (analyses/configuration, contract tooling, manufacturing plan, specification, and evaporator ammonia liquid distribution system); and F) rotating machinery (performance characteristics, radial inflow turbine; item descriptions; weight calculation-rotor; producibility analysis; long lead-time items; spares; support equipment; non recurring costs; performance characteristics-axial flow turbine; Worthington pump data; and American M.A.N. Corporation data). Also included is attachment 1 to the phase I final report which presents details of the system modeling; design, materials considerations, and systems analysis of the baseline module; system cost analysis; and supporting data. (WHK)

  19. Guidelines for calculation of atmospheric dispersion and radiological consequences of design basis reactor accidents - Severe accident calculation guidelines, EPR

    International Nuclear Information System (INIS)

    Martens, R.; Schmitz, B.M.; Horn, M.

    1999-01-01

    The activities carried out within the (reduced) project period (1. Sept. until 31. Dec. 1998) for coordinated harmonization between France and Germany, of guidelines for calculation of the radiological consequences of a severe reactor accident, are summarized. (orig./CB) [de

  20. Ocean thermal energy conversion power system development-I. Phase I. Preliminary design report. Volume 1. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-18

    The results of a conceptual and preliminary design study of Ocean Thermal Energy Conversion (OTEC) closed loop ammonia power system modules performed by Lockheed Missiles and Space Company, Inc. (LMSC) are presented. This design study is the second of 3 tasks in Phase I of the Power System Development-I Project. The Task 2 objectives were to develop: 1) conceptual designs for a 40 to 50-MW(e) closed cycle ammonia commercial plant size power module whose heat exchangers are immersed in seawater and whose ancillary equipments are in a shirt sleeve environment; preliminary designs for a modular application power system sized at 10-MW(e) whose design, construction and material selection is analogous to the 50 MW(e) module, except that titanium tubes are to be used in the heat exchangers; and 3) preliminary designs for heat exchanger test articles (evaporator and condenser) representative of the 50-MW(e) heat exchangers using aluminum alloy, suitable for seawater service, for testing on OTEC-1. The reference ocean platform was specified by DOE as a surface vessel with the heat exchanger immersed in seawater to a design depth of 0 to 20 ft measured from the top of the heat exchanger. For the 50-MW(e) module, the OTEC 400-MW(e) Plant Ship, defined in the Platform Configuration and Integration study, was used as the reference platform. System design, performance, and cost are presented. (WHK)

  1. Preliminary design concept of HYPER cooling system using Pb-Bi coolant

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang H

    2001-09-01

    The present study focuses on providing the basic concept of HYPER's cooling system based on simple and fundamental calculations. The system operating temperature was preliminarily determined as 340/510 .deg. C. The total system flow rate of HYPER is {approx} 40,000kg/sec and the flow velocity in the core is preliminarily designed to be {approx}1.5 m/sec. For hot conditions of HYPER core, the simple analytic calculation predicted that the maximum temperature of the cladding outer surface is 634 .deg. C, which is below the design limit, 650 .deg. C. However, the SLTHEN code modified for HYPER's subchannel analysis predicted that the maximum temperature of the cladding outer surface in the same conditions is higher than the design limit by 4.7 .deg. C. The comparison with the results of the analytic model and additional sensitivity calculations showed that the modified SLTHEN code can reasonably simulate the heat transfer between subchannels of the HYPER core and be used effectively for thermal hydraulic design of the HYPER core in conceptual design stage. A forced circulation is inevitable during a full power condition since natural circulation is not sufficient to cool the core with reasonable system pressure drop and reasonable system height. However, a natural circulation can be an excellent method for decay heat removal when the height difference between the core and the heat exchanger is above 10 m. In order to avoid high pressure loads on the vessel, loop configuration was chosen. The simplification of cooling system and high system efficiency were attained by removing independent target cooling system and intermediate heat transport system. A superheated rankle cycle was chosen since it is technically matured and its thermal efficiency is reasonably high.

  2. Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

    2013-01-01

    This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

  3. The neutral beam test facility cryopumping operation: preliminary analysis and design of the cryogenic system

    International Nuclear Information System (INIS)

    Gravil, B.; Henry, D.; Cordier, J.J.; Hemsworth, R.; Van Houtte, D.

    2004-01-01

    The ITER neutral beam heating and current drive system is to be equipped with a cryosorption cryopump made up of 12 panels connected in parallel, refrigerated by 4.5 K 0.4 MPa supercritical helium. The pump is submitted to a non homogeneous flux of H 2 or D 2 molecules, and the absorbed flux varies from 3 Pa.m -3 .s -1 to 35 Pa.m -3 .s -1 . In the frame of the 'ITER first injector and test facility CSU-EFDA task' (TW3-THHN-IITF1), the ITER reference cryo-system and cryo-plant designs have been assessed and compared to optimised designs devoted to the Neutral Beam Test Facility (NBTF). The 4.5 K cryo-panel, which has a mass of about 1000 kg, must be periodically regenerated up to 90 K and occasionally to 470 K. The cool-down time after regeneration depends strongly on the refrigeration capacity. Fast regeneration and cool-down of the cryo-panels are not considered a priority for the test facility operation, and an analysis of the consequences of a limited cold power refrigerator on the cooling down time has been carried out and will be discussed. This paper presents a preliminary evaluation of the NBTF cryo-plant and the associated process flow diagram. (authors)

  4. Preliminary design report for prototypical spent nuclear fuel rod consolidation equipment

    International Nuclear Information System (INIS)

    Judson, B.F.; Maillet, J.; O'Neill, G.L.; Tsitsichvili, J.; Tucoulat, D.

    1986-12-01

    The purpose of the Prototypical Consolidation Demonstration Project (PCDP) is to develop and demonstrate the equipment system that will be used to consolidate the bulk of the spent nuclear fuel generated in the United States prior to its placement in a geological repository. The equipment must thus be capable of operating on a routine production basis over a long period of time with stringent requirements for safety, reliability, productivity and cost-effectiveness. Four phases are planned for the PCDP. Phase 1 is the Preliminary Design of generic consolidation equipment that could be installed at a Monitored Retrievable Storage (MRS) facility or in the Receiving ampersand Handling Facility at a geologic repository site. Phase 2 will be the Final Design and preparation of procurement packages for the equipment in a configuration capable of being installed and tested in a special enclosure within the TAN Hot Shop at DOE's Idaho National Engineering Laboratory. In Phase 3 the equipment will be fabricated and then tested with mock fuel elements in a contractor's facility. Finally, in Phase 4 the equipment will be moved to the TAN facility for demonstration operation with irradiated spent fuel elements. 55 figs., 15 tabs

  5. Preliminary Design and Computational Fluid Dynamics Analysis of Supercritical Carbon Dioxide Turbine Blade

    International Nuclear Information System (INIS)

    Jeong, Wi S.; Kim, Tae W.; Suh, Kune Y.

    2007-01-01

    The supercritical gas turbine Brayton cycle has been adopted in the secondary loop of the Generation IV Nuclear Energy Systems, and planned to be installed in power conversion cycles of the nuclear fusion reactors as well. The supercritical carbon dioxide (SCO 2 ) is one of widely considered fluids for this concept. The potential beneficiaries include the Secure Transportable Autonomous Reactor- Liquid Metal (STAR-LM), the Korea Advanced Liquid Metal Reactor (KALIMER) and Battery Omnibus Reactor Integral System (BORIS) which is being developed at the Seoul National University. The reason for these welcomed applications is that the SCO 2 Brayton cycle can achieve higher overall energy conversion efficiency than the steam turbine Rankine cycle. Seoul National University has recently been working on the SCO 2 based Modular Optimized Brayton Integral System (MOBIS). The MOBIS design power conversion efficiency is about 45%. Gas turbine design is crucial part in achieving this high efficiency. In this paper, the preliminary analysis on first stage of gas turbine was performed using CFX as a solver

  6. Preliminary design and economic investigations of Diffuser-Augmented Wind Turbines (DAWT)

    Energy Technology Data Exchange (ETDEWEB)

    Foreman, K.M.

    1981-12-01

    A preferred design and configuration approach is suggested for the DAWT innovative wind energy conversion system. A preliminary economic asessment is made for limited production rates of units between 5 and 150 kw rated output. Nine point designs are used to arrive at the conclusions regarding best construction material for the diffuser and busbar cost of electricity (COE). It is estimated that for farm and REA cooperative end users, the COE can range between 2 and 3.5 cents/kWh for sites with annual average wind speeds of 16 and 12 mph (25.7 and 19.3 km/h) respectively, and 150 kW rated units. No tax credits are included in these COE figures. For commercial end users of these 150 kW units the COE ranges between 4.0 and 6.5 cents/kWh for 16 and 12 mph sites. These estimates in 1979 dollars are lower than DOE goals set in 1978 for the rating size and end applications. Recommendations are made for future activities to maintain steady, systematic progress toward mature development of the DAWT.

  7. Preliminary conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR). Status report, January 1978--March 1978

    International Nuclear Information System (INIS)

    Kelly, J.L.

    1978-03-01

    The DTHR preliminary conceptual design consists of a magnetically confined fusion reactor fitted with a fertile thorium blanket. The fusion driver concept is based on a beam driven plasma, but at sufficiently high plasma densities that neutrons originating from the interactions of bulk plasma ions contribute significantly to the wall loading. The tokamak has a major radius of 5.2 m, a minor radius of 1.2 m, and the elongation is 1.6. All of the magnetic coil systems are superconducting Nb 3 Sn based on the Large Coil Project (LCP) technology. The toroidal field (TF) coils employ an innovative concept, the ''compact D'' configuration. An engineered bundle divertor concept has been developed based on the bundle divertor design techniques developed for TNS and ISX-B. A thermal power of 150MW of 200 keV deuterium is injected into the plasma through six ducts of a positive ion, neutral beam injection system (NBIS). A water cooled, 316 stainless steel vacuum vessel concept was developed and initial scoping analyses look encouraging. The fusile fuel handling system was evaluated and defined. Details of the tritium injection system remain to be developed. Tritium breeding will be assessed in subsequent phases of the DTHR operation. The fusion driver provides a neutron first wall loading of 2MW/m 2 for fissile production in the blanket

  8. SolarOil Project, Phase I preliminary design report. [Solar Thermal Enhanced Oil Recovery project

    Energy Technology Data Exchange (ETDEWEB)

    Baccaglini, G.; Bass, J.; Neill, J.; Nicolayeff, V.; Openshaw, F.

    1980-03-01

    The preliminary design of the Solar Thermal Enhanced Oil Recovery (SolarOil) Plant is described in this document. This plant is designed to demonstrate that using solar thermal energy is technically feasible and economically viable in enhanced oil recovery (EOR). The SolarOil Plant uses the fixed mirror solar concentrator (FMSC) to heat high thermal capacity oil (MCS-2046) to 322/sup 0/C (611/sup 0/F). The hot fluid is pumped from a hot oil storage tank (20 min capacity) through a once-through steam generator which produces 4.8 MPa (700 psi) steam at 80% quality. The plant net output, averaged over 24 hr/day for 365 days/yr, is equivalent to that of a 2.4 MW (8.33 x 10/sup 6/ Btu/hr) oil-fired steam generator having an 86% availability. The net plant efficiency is 57.3% at equinox noon, a 30%/yr average. The plant will be demonstrated at an oilfield site near Oildale, California.

  9. Triangular SPECT system for 3-D total organ volume imaging: Design concept and preliminary imaging results

    International Nuclear Information System (INIS)

    Lim, C.B.; Anderson, J.; Covic, J.

    1985-01-01

    SPECT systems based on 2-D detectors for projection data collection and filtered back-projection image reconstruction have the potential for true 3-D imaging, providing contiguous slice images in any orientation. Anger camera-based SPECT systems have the natural advantage supporting planar imaging clinical procedures. However, current systems suffer from two drawbacks; poor utilization of emitted photons, and inadequate system design for SPECT. A SPECT system consisting of three rectangular cameras with radial translation would offer the variable cylindrical FOV of 25 cm to 40 cm diameter allowing close detector access to the object. This system would provide optimized imaging for both brain and body organs in terms of sensitivity and resolution. For brain imaging a tight detector triangle with fan beam collimation, matching detector UFOV to the head, allows full 2 π utilization of emitted photons, resulting in >4 times sensitivity increase over the single detector system. Minification of intrinsic detector resolution in fan beam collimation further improves system resolution. For body organ imaging the three detectors with parallel hole collimators, rotating in non-circular orbit, provide both improved resolution and three-fold sensitivity increase. Practical challenges lie in ensuring perfect image overlap from three detectors without resolution degradation and artifact generation in order to benefit from the above improvements. An experimental system has been developed to test the above imaging concept and we have successfully demonstrated the superior image quality of the overlapped images. Design concept will be presented with preliminary imaging results

  10. Preliminary Core Design Analysis of a 200MWth Pebble Bed-type VHTR

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Noh, Jae Man

    2007-01-01

    This paper intends to suggest the preliminary core design analysis of a VHTR for a hydrogen production. The nuclear hydrogen system that utilizes the high temperature heat generated from the VHTR is a promising candidate for a cost effective, safe and clean supply of hydrogen in the age of hydrogen economy. Among two candidate VHTR cores, that is, a prismatic modular reactor (PMR) and a pebble bed-type reactor (PBR), we focus on the design of a 200MWth PBR (hereinafter PBR200) in this paper. Here, the 200MWth power is selected for a demonstration plant. The core configuration of the PBR200 is similar to the PBMR (Pebble Bed Modular Reactor, 400MWth) of South Africa, but the overall dimension of the reactor system is scaled-down. This paper is to suggest two candidate PBR200 cores. One is an annular core with an inner reflector (PBR200-CD1) which was presented at IWRES07, and the other is a cylindrical core without an inner reflector (PBR200-CD2)

  11. New technologies and new performances of the JCMT radio-telescope: a preliminary design study

    Science.gov (United States)

    Mian, S.; De Lorenzi, S.; Ghedin, L.; Rampini, F.; Marchiori, G.; Craig, S.

    2012-09-01

    With a diameter of 15m the James Clerk Maxwell Telescope (JCMT) is the largest astronomical telescope in the world designed specifically to operate in the submillimeter wavelength region of the spectrum. It is situated close to the summit of Mauna Kea, Hawaii, at an altitude of 4092m. Its primary reflector currently consists of a steel geodesic supporting structure and pressed aluminium panels on a passive mount. The major issues of the present reflector are its thermal stability and its panels deterioration. A preliminary design study for the replacement of the JCMT antenna dish is here presented. The requested shape error for the new reflector is <20μm RMS. The proposed solution is based on a semi-monocoque backing structure made of CFRP and on high precision electroformed panels. The choice of CFRP for the backing structure allows indeed to improve the antenna performance in terms of both stiffness and thermal stability, so that the required surface accuracy of the primary can be achieved even by adopting a passive panels system. Moreover thanks to CFRP, a considerable weight reduction of the elevation structure can be attained. The performance of the proposed solution for the JCMT antenna has been investigated through FE analyses and the assessed deformation of the structure under different loading cases has been taken into account for subsequent error budgeting. Results show that the proposed solution is in line with the requested performance. With this new backing structure, the JCMT would have the largest CFRP reflector ever built.

  12. Preliminary design of a low-cost greenhouse for salt production in Indonesia

    Science.gov (United States)

    Jaziri, A. A.; Guntur; Setiawan, W.; Prihanto, A. A.; Kurniawan, A.

    2018-04-01

    Salt is an assential material of industry, not only in food industry point of view but also in various industries such as chemical, oil drilling, and animal feed industries, even less than half of salt needs used to household consumption. It is crucial to ensure salt production in Indonesia reaches the national target (3.7 million tons) due to relatively low technology and production level. Thus salt production technology is developed to facilitate farmers consisted of geomembrane and filtering-threaded technology. However, the use of those technologies in producing salt was proved less effective due to unpredictable weather conditions. Therefore, greenhouse technology is proposed to be used for salt production for several good reasons. This paper describes the preliminary design of a low-cost greenhouse designed as a pyramid model that uses bamboo, mono-layer and high density polyethylene plastics. The results confirmed that the yield of salt produced by greenhouse significantly incresed compared with prior technology and the NaCl content increased as well. The cost of greenhouse was IDR 5,688,000 and easy to assembly.

  13. Preliminary conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR). Status report, January 1978--March 1978

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.L. (ed.)

    1978-03-01

    The DTHR preliminary conceptual design consists of a magnetically confined fusion reactor fitted with a fertile thorium blanket. The fusion driver concept is based on a beam driven plasma, but at sufficiently high plasma densities that neutrons originating from the interactions of bulk plasma ions contribute significantly to the wall loading. The tokamak has a major radius of 5.2 m, a minor radius of 1.2 m, and the elongation is 1.6. All of the magnetic coil systems are superconducting Nb/sub 3/Sn based on the Large Coil Project (LCP) technology. The toroidal field (TF) coils employ an innovative concept, the ''compact D'' configuration. An engineered bundle divertor concept has been developed based on the bundle divertor design techniques developed for TNS and ISX-B. A thermal power of 150MW of 200 keV deuterium is injected into the plasma through six ducts of a positive ion, neutral beam injection system (NBIS). A water cooled, 316 stainless steel vacuum vessel concept was developed and initial scoping analyses look encouraging. The fusile fuel handling system was evaluated and defined. Details of the tritium injection system remain to be developed. Tritium breeding will be assessed in subsequent phases of the DTHR operation. The fusion driver provides a neutron first wall loading of 2MW/m/sup 2/ for fissile production in the blanket.

  14. Research on the Reform of the Preliminary Course of Architectural Design Based on Innovation & Practice Ability Training

    Science.gov (United States)

    Yuping, Cai; Shuang, Liang

    2017-01-01

    The traditional undergraduate education mode of architecture has been unable to adapt to the rapid development of society. Taking the junior professional course of architecture--the preliminary course of architectural design as an example, this paper analyzes the problems existing in the current professional courses of lower grades, puts forward…

  15. DEMONSTRATION OF FUEL CELLS TO RECOVER ENERGY FROM ANAEROBIC DIGESTER GAS - PHASE I. CONCEPTUAL DESIGN, PRELIMINARY COST, AND EVALUATION STUDY

    Science.gov (United States)

    The report discusses Phase I (a conceptual design, preliminary cost, and evaluation study) of a program to demonstrate the recovery of energy from waste methane produced by anaerobic digestion of waste water treatment sludge. The fuel cell is being used for this application becau...

  16. Grid connected integrated community energy system. Phase II: final stage 2 report. Preliminary design of cogeneration plant

    Energy Technology Data Exchange (ETDEWEB)

    1978-03-22

    The preliminary design of a dual-purpose power plant to be located on the University of Minnesota is described. This coal-fired plant will produce steam and electric power for a grid-connected Integrated Community Energy System. (LCL)

  17. A fast-track preliminary thermo-mechanical design of oil export pipelines from P-56 platform

    Energy Technology Data Exchange (ETDEWEB)

    Solano, Rafael F.; Mendonca, Salete M. de [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil); Franco, Luciano D.; Walker, Alastair; El-Gebaly, Sherif H. [INTECSEA, Rio de Janeiro, RJ (Brazil)

    2009-12-19

    The oil export pipelines of Marlim Sul field Module 3, Campus Basin, offshore Brazil, will operate in high pressure and temperature conditions, and will be laid on seabed crossing ten previously laid pipelines along the routes. In terms of thermo-mechanical design, these conditions turn out to be great challenges. In order to obtain initial results and recommendations for detail design, a preliminary thermo-mechanical design of pipelines was carried out as a fast-track design before the bid. This way, PETROBRAS can assess and emphasize the susceptibility of these lines to lateral buckling and pipeline walking behavior. Therefore, PETROBRAS can present a preliminary mitigation strategy for lateral buckling showing solutions based on displacement controlled criteria and by introducing buckle initiation along the pipeline using distribution buoyancy. Besides that, axial displacements and loads at the pipeline ends can be furnished also in order to provide a basis for the detailed design. The work reported in this paper follows the SAFEBUCK JIP methodology and recommendation, which were used to determine the allowable strain and maximum allowable VAS (Virtual Anchor Spacing) considered in the buckling mitigation strategy. The paper presents also the formation of uncontrolled buckles on the seabed and the propensity for pipeline walking in its sections between buckles. The buckling mitigation strategy established in this preliminary design confirms that the oil pipeline specifications are adequate to maintain integrity during design life. (author)

  18. Preliminary design study for a carbide LEU-nuclear thermal rocket

    International Nuclear Information System (INIS)

    Venneri, P.F.; Kim, Y.

    2014-01-01

    Nuclear space propulsion is a requirement for the successful exploration of the solar system. It offers the possibility of having both a high specific impulse and a relatively high thrust, allowing rapid transit times with a minimum usage of fuel. This paper proposes a nuclear thermal rocket design based on heritage NERVA rockets that makes use of Low Enriched Uranium (LEU) fuel. The Carbide LEU Nuclear Thermal Rocket (C-LEU-NTR) is designed to fulfill the rocket requirements as set forth in the NASA 2009 Mars Mission Design Reference Architecture 5.0, that is provide 25,000 lbf of thrust, operate at full power condition for at least two hours, and have a specific impulse close to 900 s. The neutronics analysis was done using MCNP5 with the ENDF/B-VII.1 neutron library. The thermal hydraulic calculations and size optimization were completed with a finite difference code being developed at the Center for Space Nuclear Research. (authors)

  19. Optics calculations and beam line design for the JANNuS facility in Orsay

    International Nuclear Information System (INIS)

    Chauvin, N.; Henry, S.; Flocard, H.; Fortuna, F.; Kaitasov, O.; Pariset, P.; Pellegrino, S.; Ruault, M.O.; Serruys, Y.; Trocelier, P.

    2007-01-01

    JANNuS (Joint Accelerators for Nano-Science and Nuclear Simulation) will be a unique user facility in Europe dedicated to material modification by ion beam implantation and irradiation. The main originality of the project is that it will be possible to perform implantation and irradiation with simultaneous multiple ions beams and in situ characterization by transmission electron microscopy (TEM) observation or ion beam analysis. This facility will be composed of two experimental platforms located in two sites: the CEA-SRMP in Saclay and the CNRS-CSNSM in Orsay. This paper will focus on the design of two new transport beam lines for the Orsay site. One of the most challenging parts of the JANNuS project (Orsay site) is to design two new beam lines in order to inject, into a 200 kV TEM, two different ion beams (low and medium energy) coming from two existing pieces of equipment: a 2 MV Tandem accelerator and a 190 kV ion implanter. For these new beam lines, first order beam calculations have been done using transfer matrix formalism. A genetic algorithm has been written and adapted to perform the optimization of the beam line parameters. Then, using the SIMION code, field maps of the electrostatic elements (quadrupoles, spherical sectors) have been calculated and ion trajectories have been simulated. We studied specifically the optical aberrations induced by the electrostatic spherical deflectors. Finally, the results of the first order calculations and the field map simulations show a good agreement

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-06-01

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.