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Sample records for demonstration plant design

  1. US GCFR demonstration plant design

    International Nuclear Information System (INIS)

    Hunt, P.S.; Snyder, H.J.

    1980-05-01

    A general description of the US GCFR demonstration plant conceptual design is given to provide a context for more detailed papers to follow. The parameters selected for use in the design are presented and the basis for parameter selection is discussed. Nuclear steam supply system (NSSS) and balance of plant (BOP) component arrangements and systems are briefly discussed

  2. Design and simulation of a plant control system for a GCFR demonstration plant

    International Nuclear Information System (INIS)

    Estrine, E.A.; Greiner, H.G.

    1980-02-01

    A plant control system is being designed for a 300 MW(e) Gas Cooled Fast Breeder Reactor (GCFR) demonstration plant. Control analysis is being performed as an integral part of the plant design process to ensure that control requirements are satisfied as the plant design evolves. Plant models and simulations are being developed to generate information necessary to further define control system requirements for subsequent plant design iterations

  3. GCFR demonstration plant: conceptual design and status report

    International Nuclear Information System (INIS)

    1980-12-01

    Helium Breeder Associates (HBA), a non-profit corporation, has been the program manager and technical integrator of the Gas-Cooled Fast Reactor (GCFR) development effort since 1977. When DOE discontinued support of the GCFR in 1980, the HBA members undertook the task of providing for an orderly termination and documentation of the program. HBA does not agree with the government's rational for withdrawing support for this promising technology and has directed its termination and documentation toward preserving the current state of its development. Toward that end, HBA has compiled the following report which is a summary of the conceptual design of the demonstration plant and status of the program as of the end of 1980. It includes summaries of tasks that have not evolved to a final conclusion. Although the report has not been subjected to formal review and approval by the designers, it is intended to provide the reader with the design considerations that were current at the time of program termination. It is hoped that the report will be useful in restarting the program in the future by establishing the basis of the completed conceptual design and indicating a logical path for new design and development

  4. Phase I: the pipeline-gas demonstration plant. Demonstration plant engineering and design. Volume 18. Plant Section 2700 - Waste Water Treatment

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-05-01

    Contract No. EF-77-C-01-2542 between Conoco Inc. and the US Department of Energy provides for the design, construction, and operation of a demonstration plant capable of processing bituminous caking coals into clean pipeline quality gas. The project is currently in the design phase (Phase I). This phase is scheduled to be completed in June 1981. One of the major efforts of Phase I is the process and project engineering design of the Demonstration Plant. The design has been completed and is being reported in 24 volumes. This is Volume 18 which reports the design of Plant Section 2700 - Waste Water Treatment. The objective of the Waste Water Treatment system is to collect and treat all plant liquid effluent streams. The system is designed to permit recycle and reuse of the treated waste water. Plant Section 2700 is composed of primary, secondary, and tertiary waste water treatment methods plus an evaporation system which eliminates liquid discharge from the plant. The Waste Water Treatment Section is designed to produce 130 pounds per hour of sludge that is buried in a landfill on the plant site. The evaporated water is condensed and provides a portion of the make-up water to Plant Section 2400 - Cooling Water.

  5. Over facility design description for the CPDF [Centrifuge Plant Demonstration Facility]: SDD-1 [System Design Description

    International Nuclear Information System (INIS)

    1987-04-01

    The Centrifuge Plant Demonstration Facility (CPDF) is an essential part of the continuing development of first-production-plant centrifuge technology that will integrate centrifuge machines into a process and enrichment plant design. The CPDF will provide facilities for testing and continued development of a unit cascade in direct support of the commercial Gas Centrifuge Enrichment Plant (GCEP). The basic cascade-oriented equipment, feed, withdrawal, drive system, process piping, utility piping, and other auxiliary and support equipment will be tested in an operating configuration that represents, to the extent possible, GCEP arrangement and operating conditions. The objective will be to demonstrate procedures for production cascade installation, start-up, operation, and maintenance, and to provide proof of overall cascade and associated system design, construction, and operating and maintenance concepts. To the maximum possible extent, all equipment for the CPDF will be procured from commercial sources. Centrifuges will be procured from industry using government-supplied specifications and drawings. The existing Component Preparation Laboratory (CPL) located near the CPDF site will be used for centrifuge component receiving, inspection, assembly, and qualification testing of pre-production test machines. Later in the test program, samples of production machines planned for use in the GCEP will be tested in the CPDF

  6. Demonstration tokamak power plant

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

  7. Design consideration on hydrogen production demonstration plant of thermochemical IS process

    International Nuclear Information System (INIS)

    Iwatsuki, Jin; Noguchi, Hiroki; Terada, Atsuhiko; Kubo, Shinji; Sakaba, Nariaki; Onuki, Kaoru; Hino, Ryutaro

    2009-03-01

    Preliminary design study was carried out on the hydrogen production demonstration plant of thermochemical IS process. In the pilot test, hydrogen production will be examined under prototypical condition using an apparatus made of industrial materials, which is driven by the sensible heat of helium gas heated by an electric heater that simulates the High Temperature Engineering Test Reactor (HTTR). Tentative system condition was defined considering the HTTR specification and the experience on the construction and the operation of the mock-up test facility using methane reforming for hydrogen production. The process condition and the system flow diagram were discussed to meet the system condition. Based on the defined process condition, types of the main components were discussed taking the corrosion resistance of the structural materials into consideration. Applicable rules and regulations were also surveyed regarding the plant construction and operation. (author)

  8. GCFR demonstration plant: conceptual design and status report. Briefing document and executive summary

    International Nuclear Information System (INIS)

    1980-12-01

    Helium Breeder Associates (HBA), a non-profit corporation, has been the program manager and technical integrator of the Gas-Cooled Fast Reactor (GCFR) development effort since 1977. When DOE discontinued support of the GCFR in 1980, the HBA members undertook the task of providing for an orderly termination and documentation of the program. HBA does not agree with the government's rational for withdrawing support for this promising technology and has directed its termination and documentation toward preserving the current state of its development. Toward that end, HBA has compiled a report of which this is a brief summary covering the conceptual design of the demonstration plant and status of the program as of the end of 1980. All the work in design, development, safety, and licensing which has been done to date, both in this country and in Europe, has given positive results and the prognosis for ultimately achieving the potential advantages of the GCFR is excellent. The report is intended to provide the reader with the design considerations that were current at the time of program termination. The report could be useful in restarting the program in the future by establishing the basis of the completed conceptual design and indicating a logical path for new design and development

  9. Accelerator production of tritium plant design and supporting engineering development and demonstration work

    International Nuclear Information System (INIS)

    Lisowski, P.W.

    1997-11-01

    Tritium is an isotope of hydrogen with a half life of 12.3 years. Because it is essential for US thermonuclear weapons to function, tritium must be periodically replenished. Since K reactor at Savannah River Site stopped operating in 1988, tritium has been recycled from dismantled nuclear weapons. This process is possible only as long as many weapons are being retired. Maintaining the stockpile at the level called for in the present Strategic Arms Reduction Treaty (START-I) will require the Department of Energy to have an operational tritium production capability in the 2005--2007 time frame. To make the required amount of tritium using an accelerator based system (APT), neutrons will be produced through high energy proton reactions with tungsten and lead. Those neutrons will be moderated and captured in 3 He to make tritium. The APT plant design will use a 1,700 MeV linear accelerator operated at 100 mA. In preparation for engineering design, starting in October 1997 and subsequent construction, a program of engineering development and demonstration is underway. That work includes assembly and testing of the first 20 MeV of the low energy plant linac at 100 mA, high-energy linac accelerating structure prototyping, radiofrequency power system improvements, neutronic efficiency measurements, and materials qualifications

  10. Fuel Gas Demonstration Plant Program. Volume I. Demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    The objective of this project is for Babcock Contractors Inc. (BCI) to provide process designs, and gasifier retort design for a fuel gas demonstration plant for Erie Mining Company at Hoyt Lake, Minnesota. The fuel gas produced will be used to supplement natural gas and fuel oil for iron ore pellet induration. The fuel gas demonstration plant will consist of five stirred, two-stage fixed-bed gasifier retorts capable of handling caking and non-caking coals, and provisions for the installation of a sixth retort. The process and unit design has been based on operation with caking coals; however, the retorts have been designed for easy conversion to handle non-caking coals. The demonstration unit has been designed to provide for expansion to a commercial plant (described in Commercial Plant Package) in an economical manner.

  11. Design features of a reverse osmosis demonstration plant for treatment of low level radioactive waste

    International Nuclear Information System (INIS)

    Shekhar, P.; Sudesh Nath; Gandhi, P.M.; Mishra, S.D.

    1994-01-01

    Reverse osmosis, a novel process in the field of nuclear waste management, is under evaluation globally. Its application is basically considered for the treatment of low level waste; yet references are found for its possible use to treat specific intermediate level waste streams, if segregated at source. The process of reverse osmosis (RO) is proposed for use in conjunction with other conventional processes like chemical treatment, ion exchange and evaporation. Flow sheets have been developed wherein RO can come as a replacement of one of these processes or is used as a pre or post treatment stage. The emphasis is on reducing the secondary wastes so as to realize an optimum levelised cost of treatment. This paper outlines the design basis for an RO plant for treating low level radioactive wastes based on the studies carried out on laboratory as well as bench scale. (author)

  12. Design features of a reverse osmosis demonstration plant for treatment of low level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Shekhar, P; Nath, Sudesh; Gandhi, P M; Mishra, S D [Waste Management Projects Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Reverse osmosis, a novel process in the field of nuclear waste management, is under evaluation globally. Its application is basically considered for the treatment of low level waste; yet references are found for its possible use to treat specific intermediate level waste streams, if segregated at source. The process of reverse osmosis (RO) is proposed for use in conjunction with other conventional processes like chemical treatment, ion exchange and evaporation. Flow sheets have been developed wherein RO can come as a replacement of one of these processes or is used as a pre or post treatment stage. The emphasis is on reducing the secondary wastes so as to realize an optimum levelised cost of treatment. This paper outlines the design basis for an RO plant for treating low level radioactive wastes based on the studies carried out on laboratory as well as bench scale. (author). 3 figs., 3 tabs.

  13. The design and operation of a continuous ion-exchange demonstration plant for the recovery of uranium

    International Nuclear Information System (INIS)

    Craig, W.M.; Douglas, M.E.E.; Louw, G.D.

    1978-01-01

    A description is given of the design of the continuous ion-exchange demonstration plant at Blyvooruitzicht Gold Mine, including details of the process design, the column construction, and the control system. The operating and process results gathered over a period of seventeen months are summarized, and devolopment work and modifications to the process are discussed. It is concluded that the system comprising continuous loading and continuous elution is technically feasible and can be scaled up with confidence [af

  14. Design of a hot pilot plant facility for demonstration of the pot calcination process

    Energy Technology Data Exchange (ETDEWEB)

    Buckham, J A

    1962-01-01

    A facility was designed for demonstration of the pot calcination process with wastes from processing aluminum alloy fuels, Darex or electrolytic processing of stainless-steel fuels, and Purex processes. This facility will also permit determination of procedures required for economical production of low-porosity, relatively nonleachable materials by addition of suitable reagents to the wastes fed to the calciner. The process consists of concentration by evaporation and thermal decomposition in situ in pots which also serve as the final disposal containers. This unit permits determination of pot loading and density, leachability, melting point, volatile material content, heat release, and thermal conductivity of the calcine. Also to be determined are transient calcine temperature distributions, fission product behavior during calcination, deentrainment obtained in the various parts of the system, decontamination achieved on all liquid and gaseous effluent streams, need for venting of stored pots, optimum means of remotely sealing the pots, and methods required for production of a minimum volume of noncondensable off-gas. This facility will employ nominal full-scale pots 8 and 12 in. in diameter and 8 ft long. A unique evaporator design was evolved to permit operation either with close-coupled continuous feed preparation or with bath feed preparation. Provisions were made to circumvent possible explosions due to organic material in feed solutions and other suspected hazards.

  15. Industrial Fuel Gas Demonstration Plant Program. Conceptual design and evaluation of commercial plant. Volume III. Economic analyses (Deliverable Nos. 15 and 16)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-01-01

    This report presents the results of Task I of Phase I in the form of a Conceptual Design and Evaluation of Commercial Plant report. The report is presented in four volumes as follows: I - Executive Summary, II - Commercial Plant Design, III - Economic Analyses, IV - Demonstration Plant Recommendations. Volume III presents the economic analyses for the commercial plant and the supporting data. General cost and financing factors used in the analyses are tabulated. Three financing modes are considered. The product gas cost calculation procedure is identified and appendices present computer inputs and sample computer outputs for the MLGW, Utility, and Industry Base Cases. The results of the base case cost analyses for plant fenceline gas costs are as follows: Municipal Utility, (e.g. MLGW), $3.76/MM Btu; Investor Owned Utility, (25% equity), $4.48/MM Btu; and Investor Case, (100% equity), $5.21/MM Btu. The results of 47 IFG product cost sensitivity cases involving a dozen sensitivity variables are presented. Plant half size, coal cost, plant investment, and return on equity (industrial) are the most important sensitivity variables. Volume III also presents a summary discussion of the socioeconomic impact of the plant and a discussion of possible commercial incentives for development of IFG plants.

  16. AAEC builds synroc demonstration plant

    International Nuclear Information System (INIS)

    O'Hagan, R.

    1986-01-01

    A demonstration plant to test the feasibility of an Australian-developed method of immobilising radioactive waste is being built at the Australian Atomic Energy Commission's Lucas Heights Research Laboratories. The plant will operate as if radioactive waste was actually being processed, but non-radioactive elements of a similar composition will be used. The process involves the simulated waste being mixed into a slurry with the main SYNROC ingredients and then converted to a powder. The powder is moved about the plant in bellows-type containers by robots

  17. Savannah River Plant incinerator demonstration

    International Nuclear Information System (INIS)

    Lewandowski, K.E.

    1983-01-01

    A full-scale incineration process was demonstrated at the Savannah River Laboratory (SRL) using nonradioactive waste. From October 1981 through September 1982, 15,700 kilograms of solid waste and 5.7 m 3 of solvent were incinerated. Emissions of off-gas components (NO/sub x/, SO 2 , CO, and particulates) were well below South Carolina state standards. Volume reductions of 20:1 for solid waste and 7:1 for Purex solvent/lime slurry were achieved. The process has been relocated and upgraded by the Savannah River Plant to accept low-level beta-gamma combustibles. During a two-year demonstration, the facility will incinerate slightly radioactive ( 3 ) solvent and suspect level (< 1 mR/h at 0.0254 meter) solid wastes. This demonstration will begin in early 1984

  18. Flambeau River Biofuels Demonstration Plant

    Energy Technology Data Exchange (ETDEWEB)

    Byrne, Robert J. [Flambeau River Biofuels, Inc., Park Falls, WI (United States)

    2012-07-30

    Flambeau River BioFuels, Inc. (FRB) proposed to construct a demonstration biomass-to-liquids (BTL) biorefinery in Park Falls, Wisconsin. The biorefinery was to be co-located at the existing pulp and paper mill, Flambeau River Papers, and when in full operation would both generate renewable energy – making Flambeau River Papers the first pulp and paper mill in North America to be nearly fossil fuel free – and produce liquid fuels from abundant and renewable lignocellulosic biomass. The biorefinery would serve to validate the thermochemical pathway and economic models for BTL production using forest residuals and wood waste, providing a basis for proliferating BTL conversion technologies throughout the United States. It was a project goal to create a compelling new business model for the pulp and paper industry, and support the nation’s goal for increasing renewable fuels production and reducing its dependence on foreign oil. FRB planned to replicate this facility at other paper mills after this first demonstration scale plant was operational and had proven technical and economic feasibility.

  19. Utilities and offsites design baseline. Outside Battery Limits Facility 6000 tpd SRC-I Demonstration Plant. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-05-25

    Volume 2 contains flowsheets and equipment specifications for the following parts of the plant: cooling water systems, process water supply, potable water supply, nitrogen system, compressed air system, flares, incinerators, fuels and interconnecting systems (pipes). The instrumentation requirements are included. (LTN)

  20. Utilities and offsites design baseline. Outside Battery Limits Facility 6000 tpd SRC-I Demonstration Plant. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-05-25

    As part of the overall Solvent Refined Coal (SRC-1) project baseline being prepared by International Coal Refining Company (ICRC), the RUST Engineering Company is providing necessary input for the Outside Battery Limits (OSBL) Facilities. The project baseline is comprised of: design baseline - technical definition of work; schedule baseline - detailed and management level 1 schedules; and cost baseline - estimates and cost/manpower plan. The design baseline (technical definition) for the OSBL Facilities has been completed and is presented in Volumes I, II, III, IV, V and VI. The OSBL technical definition is based on, and compatible with, the ICRC defined statement of work, design basis memorandum, master project procedures, process and mechanical design criteria, and baseline guidance documents. The design basis memorandum is included in Paragraph 1.3 of Volume I. The baseline design data is presented in 6 volumes. Volume I contains the introduction section and utility systems data through steam and feedwater. Volume II continues with utility systems data through fuel system, and contains the interconnecting systems and utility system integration information. Volume III contains the offsites data through water and waste treatment. Volume IV continues with offsites data, including site development and buildings, and contains raw materials and product handling and storage information. Volume V contains wastewater treatment and solid wastes landfill systems developed by Catalytic, Inc. to supplement the information contained in Volume III. Volume VI contains proprietary information of Resources Conservation Company related to the evaporator/crystallizer system of the wastewater treatment area.

  1. Demonstration tokamak-power-plant study (DEMO)

    International Nuclear Information System (INIS)

    1982-09-01

    A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

  2. Coal demonstration plants. Quarterly report, July--September 1977

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-02-01

    The objective of DOE's demonstration plant program is to establish the technical and financial feasibility of coal conversion technologies proven during pilot plant testing. Demonstration plants will minimize the technical and economic risks of commercialization by providing a near commercial size plant for testing and production. Thus, DOE is sponsoring the development of a series of demonstration plants, each of which will be a smaller version of commercial plants envisioned for the 1980's. These plants will be wholly integrated, self-sufficient in terms of heat generation, and dependent only on feedstock of coal, water, and air. Contracts for designing, constructing, and operating the demonstration plants will be awarded through competitive procedures and will be jointly funded. The conceptual design phase will be funded by the government, with the detailed design, procurement, construction, and operation phases being co-funded, 50% from industry and 50% from the government. The cost involved in building and operating a demonstration plant will probably be between $200 million and $500 million, depending on the size of the plant. Twenty-two projects involving demonstration plants or support projects for such plants are reviewed, including a summary for each of progress in the quarter. (LTN)

  3. Coal demonstration plants. Quarterly report, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-01-01

    The objective of DOE's demonstration plant program is to establish the technical and financial feasibility of coal conversion technologies proven during pilot plant testing. Demonstration plants will minimize the technical and economic risks of commercialization by providing a near commercial size plant for testing and production. Thus, DOE is sponsoring the development of a series of demonstration plants, each of which will be a smaller version of commercial plants envisioned for the 1980's. These plants will be wholly integrated, self-sufficient in terms of heat generation, and dependent only on feedstock of coal, water, and air. Under the DOE program, contracts for designing, constructing, and operating the demonstration plants will be awarded through competitive procedures and will be jointly funded. The conceptual design phase will be funded by the government, with the detailed design, procurement, construction, and operation phases being co-funded, 50% from industry and 50% from the government. The cost involved in building and operating a demonstration plant will probably be between $200 million and $500 million, depending on the size of the plant. Six of these demonstration plant projects are described and progress in the quarter is summarized. Several support and complementary projects are described (fuel feeding system development, performance testing and comparative evaluation, engineering support, coal grinding equipment development and a critical components test facility). (LTN)

  4. GFR demonstrator ALLEGRO design status

    International Nuclear Information System (INIS)

    Poette, C.; Malo, J.Y.; Brun-Magaud, V.; Morin, F.; Dor, I.; Mathieu, B.; Duhamel, H.; Stainsby, R.; Mikityuk, K.

    2009-01-01

    The ALLEGRO project has the ambitious goal of building and operating the first Gas Cooled Fast Reactor (GFR). It will be a low power experimental reactor with the main objective to validate on a pilot scale the specific GFR technologies (fuel element and sub-assembly, safety systems). It is a loop type, non electricity generating reactor. Its power is about 80 MW. The approach for the core includes first MOX cores loaded with some ceramic mixed carbide or nitride sub-assemblies with SiC/SiCf cladding and wrappers. When such unit test will be considered convincing enough, the diagrid and circuits are designed to accept full high temperature ceramic cores. The core neutrons can also be used to irradiate structural materials with fast neutron spectrum and in a large temperature range. The core can also include innovative irradiation fuel devices (samples or full bundles) for other reactor systems. Finally, the primary circuit can be connected to a test loop to validate the reactor coupled operation of a high temperature process or component. The paper deals with the current ALLEGRO design studies on a mid term roadmap aiming at ending the viability phase in 2012 in order to make a decision in 2013 for further detailed design and construction. Since 2005, the ALLEGRO design studies are shared in the GCFR 6th Framework Program which gathers 10 partners from 6 European countries. The paper will give an overview of recent progresses in various areas such as: - Last 3D core physics analysis of the MOX cores and their irradiation performances in terms of fast flux, dose/burnup, irradiation locations. - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the MOX core. - Fuel handling principles and solutions. - System design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy (DHR) for depressurized accidents. - An overview of the system transient analysis performed by the partners

  5. Environmental analysis for pipeline gas demonstration plants

    Energy Technology Data Exchange (ETDEWEB)

    Stinton, L.H.

    1978-09-01

    The Department of Energy (DOE) has implemented programs for encouraging the development and commercialization of coal-related technologies, which include coal gasification demonstration-scale activities. In support of commercialization activities the Environmental Analysis for Pipeline Gas Demonstration Plants has been prepared as a reference document to be used in evaluating potential environmental and socioeconomic effects from construction and operation of site- and process-specific projects. Effluents and associated impacts are identified for six coal gasification processes at three contrasting settings. In general, impacts from construction of a high-Btu gas demonstration plant are similar to those caused by the construction of any chemical plant of similar size. The operation of a high-Btu gas demonstration plant, however, has several unique aspects that differentiate it from other chemical plants. Offsite development (surface mining) and disposal of large quantities of waste solids constitute important sources of potential impact. In addition, air emissions require monitoring for trace metals, polycyclic aromatic hydrocarbons, phenols, and other emissions. Potential biological impacts from long-term exposure to these emissions are unknown, and additional research and data analysis may be necessary to determine such effects. Possible effects of pollutants on vegetation and human populations are discussed. The occurrence of chemical contaminants in liquid effluents and the bioaccumulation of these contaminants in aquatic organisms may lead to adverse ecological impact. Socioeconomic impacts are similar to those from a chemical plant of equivalent size and are summarized and contrasted for the three surrogate sites.

  6. Design of chemical plant

    International Nuclear Information System (INIS)

    Lee, Dong Il; Kim, Seung Jae; Yang, Jae Ho; Ryu, Hwa Won

    1993-01-01

    This book describes design of chemical plant, which includes chemical engineer and plan for chemical plant, development of chemical process, cost engineering pattern, design and process development, general plant construction plan, project engineering, foundation for economy on assets and depreciation, estimation for cost on capital investment and manufacturing cost, design with computers optimal design and method like fluid mechanics design chemical device and estimation for cost, such as dispatch of material and device writing on design report and appendixes.

  7. Early Site Permit Demonstration Program: Plant parameters envelope report

    International Nuclear Information System (INIS)

    1993-03-01

    The Early Site Permit (ESP) Demonstration Program is the nuclear industry's initiative for piloting the early resolution of siting-related issues before the detailed design proceedings of the combined operating license review. The ESP Demonstration Program consists of three phases. The plant parameters envelopes task is part of Phase 1, which addresses the generic review of applicable federal regulations and develops criteria for safety and environmental assessment of potential sites. The plant parameters envelopes identify parameters that characterize the interface between an ALWR design and a potential site, and quantify the interface through values selected from the Utility Requirements Documents, vendor design information, or engineering assessments. When augmented with site-specific information, the plant parameters envelopes provide sufficient information to allow ESPs to be granted based on individual ALWR design information or enveloping design information for the evolutionary, passive, or generic ALWR plants. This document is expected to become a living document when used by future applicants

  8. Desalination demonstration plant using nuclear heat

    International Nuclear Information System (INIS)

    Hanra, M.S.; Misra, B.M.

    1998-01-01

    Most of the desalination plants which are operating throughout the world utilize the energy from thermal power station which has the main disadvantage of polluting the environment due to combustion of fossil fuel and with the inevitable rise in prices of fossil fuel, nuclear driven desalination plants will become more economical. So it is proposed to set up nuclear desalination demonstration plant at the location of Madras Atomic Power Station (MAPS), Kalpakkam. The desalination plant will be of a capacity 6300 m 3 /day and based on both Multi Stage Flash (MSF) and Sea Water Reverse Osmosis (SWRO) processes. The MSF plant with performance ratio of 9 will produce water total dissolved solids (TDS-25 ppm) at a rate of 4500 m 3 /day from seawater of 35000 ppm. A part of this water namely 1000 m 3 /day will be used as Demineralised (DM) water after passing it through a mixed bed polishing unit. The remaining 3500 m 3 /day water will be mixed with 1800 m 3 /day water produced from the SWRO plant of TDS of 400 ppm and the same be supplied to industrial/municipal use. The sea water required for MSF and SWRO plants will be drawn from the intake/outfall system of MAPS which will also supply the required electric power pumping. There will be net 4 MW loss of power of MAPS namely 3 MW for MSF and 1 MW for SWRO desalination plants. The salient features of the project as well as the technical details of the both MSF and SWRO processes and its present status are given in this paper. It also contains comparative cost parameters of water produced by both processes. (author)

  9. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Tobias

    2013-07-03

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  10. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    International Nuclear Information System (INIS)

    Hartmann, Tobias

    2013-01-01

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  11. Coal demonstration plants. Quarterly report, April-June 1979

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    The objective of the US DOE demonstration program is to demonstrate and verify second-generation technologies and validate the economic, environmental and productive capacity of a near commercial-size plant by integrating and operating a modular unit using commercial size equipment. These facilities are the final stage in the RD and D process aimed at accelerating and reducing the risks of industrial process implementation. Under the DOE program, contracts for the design, construction, and operation of the demonstration plants are awarded through competitive procedures and are cost shared with the industrial partner. The conceptual design phase is funded by the government, with the detailed design, procurement, construction, and operation phases being co-funded between industry and the government. The government share of the cost involved for a demonstration plant depends on the plant size, location, and the desirability and risk of the process to be demonstrated. The various plants and programs are discussed: Description and status, funding, history, flowsheet and progress during the current quarter. (LTN)

  12. Kimberlina: a zero-emissions demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Pronske, K. [Clean Energy Systems Inc. (USA)

    2007-06-15

    FutureGen may be getting the headlines, but it is not the only superclean demonstration plant in town. In fact, you could argue that other technologies are further down the evolutionary timeline. Case in point: Clean Energy Systems' adaptation of rocket engine technology to radically change the way fuel is burned. The result is a true zero-emissions power plant. Its most distinctive element is an oxy-combustor, similar to one used in rocket engines, that generates steam by burning clean, gaseous fuel in the presence of gaseous oxygen and water. The clean fuel is prepared by processing a conventional fossil fuel such as coal-derived syngas, refinery residues, biomass or biodigester gas, or natural or landfill gas. Combustion takes place at near-stoichiometric conditions to produce a mixture of steam and CO{sub 2} at high temperature and pressure. The steam conditions are suitable for driving a conventional or advanced steam turbine-generator, or a gas turbine modified to be driven by high-temperature steam or to do work as an expansion unit at intermediate pressure. After pressure through the turbine(s), the steam/CO{sub 2} mixture is condensed, cooled, and separated into water and CO{sub 2}. The CO{sub 2} can be sequestered and/or purified and sold for commercial use. Durability and performance tests carried out between March 2005 and March 2006 produced excellent results. CO and NOx emissions are considerably low than those of combined-cycle power plants fuelled by natural gas and using selective catalytic reduction for NOx control. Work is continuing under an NETL grant. Progress and plans are reported in the article. 7 figs.

  13. Industrial Fuel Gas Demonstration Plant Program. Demonstration plant operation plan (Deliverable No. 38)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-01-01

    The Demo Plant Operating Plan is composed of the following sequence of events starting with the training or personnel, familiarizing of the personnel with the plant and completing the long-term run in the following sequences: inspection during construction, plant completion, shakedown of equipment, process unit startup, shakedown of process units, variable run operation and a turnaround. During the construction period, technical personnel from DRC, MLGW and IGT will be at the plant site becoming familiar with the equipment, its installation and all of the auxiliaries so that on completion of construction they will be well grounded on the plant detail and its configuration. At the same time the supervisory operating personnel will have hands on training the gasifier operation at the IGT pilot plant to develop a field for gasifier operation. As a plant sections are completed, they will be checked out in accordance with the contractor and operator (client) procedure as outlined. Subsequent to this, various vendor designs and furnished equipment will be checked out operating-wise and a performance test run if feasible. The actual startup of the plant will be subsequential with the support areas as utilities, coal handling and waste treatment being placed in operation first. Subsequent to this the process units will be placed in operation starting from the rear of the process train and working forward. Thus the downstream units will be operating before the reactor is run on coal. The reactor will be checked out on coke operation.

  14. The UK commercial demonstration fast reactor design

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1987-01-01

    The paper on the UK Commercial Demonstration Fast Reactor design was presented to the seminar on 'European Commercial Fast Reactor Programme, London 1987. The design is discussed under the topic headings:- primary circuit, intermediate heat exchangers and pumps, fuel and core, refuelling, steam generators, and nuclear island layout. (U.K.)

  15. Design criteria for Reedy Creek Demonstration Project

    International Nuclear Information System (INIS)

    Felicione, F.S.; Logan, J.A.

    1980-11-01

    This document defines the basic criteria for the 100-ton/day pilot plant which will use the Andco-Torrax pyrolysis process at Walt Disney World in Orlando, Florida, to produce hot water. The waste will simulate transuranic wastes which are stored at INEL. The Andco-Torrax process is designed to convert mixed municipal refuse into energy and is called slagging pyrolysis solid waste conversion

  16. Coal-fired CCS demonstration plants, 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-10-15

    The present report reviews activities taking place focused on the eventual large-scale deployment of carbon capture systems on coal-fired power plants. With this aim in mind, there are three main CO2 capture technology streams currently being developed and tested; these comprise pre-combustion capture, post-combustion capture, and systems based on oxyfuel technology. Although numerous other capture systems have been proposed, these three are currently the focus of most RD&D efforts and this report concentrates on these. More speculative technologies still at early stages in their development are not addressed. The overall aims of this report are to provide an update of recent technological developments in each of the main categories of CO2 capture, and to review the current state of development of each, primarily through an examination of larger-scale development activities taking place or proposed. However, where appropriate, data generated by smaller-scale testing is noted, especially where this is feeding directly into ongoing programmes aimed at developing further, or scaling-up the particular technology. Each is reviewed and the status of individual coal-based projects and proposals described. These are limited mainly to what are generally described as pilot and/or demonstration scale. Where available, learning experiences and operational data being generated by these projects is noted. Technology Readiness Levels (TRLs) of individual projects have been used to provide an indication of technology scale and maturity. For pre-combustion capture, post-combustion capture and oxyfuel systems, an attempt has been made to identify the technological challenges and gaps in the knowledge that remain, and to determine what technology developers are doing in terms of RD&D to address these. However, issues of commercial confidentiality have meant that in some cases, information in the public domain is limited, hence it has only been possible to identify overarching

  17. Demonstrating the Effects of Light Quality on Plant Growth.

    Science.gov (United States)

    Whitesell, J. H.; Garcia, Maria

    1977-01-01

    Describes a lab demonstration that illustrates the effect of different colors or wavelengths of visible light on plant growth and development. This demonstration is appropriate for use in college biology, botany, or plant physiology courses. (HM)

  18. Early Site Permit Demonstration Program: Station design alternatives report

    International Nuclear Information System (INIS)

    1993-03-01

    This report provides the results of investigating the basis for including Station Design Alternatives (SDAs) in the regulatory guidance given for nuclear plant environmental reports (ERs), explains approaches or processes for evaluating SDAs at the early site permit (ESP) stage, and applies one of the processes to each of the ten systems or subsystems considered as SDAS. The key objective o this report s to demonstrate an adequate examination of alternatives can be performed without the extensive development f design data. The report discusses the Composite Suitability Approach and the Established Cutoff Approach in evaluating station design alternatives and selects one of these approaches to evaluate alternatives for each of the plant or station that were considered. Four types of ALWRs have been considered due to the availability of extensive plant data: System 80+, AP600, Advanced Boiling Reactor (ABWR), and Simplified Boiling Water Reactor (SBWR). This report demonstrates the feasibility of evaluating station design alternatives when reactor design detail has not been determined, quantitatively compares the potential ental impacts of alternatives, and focuses the ultimate selection of a alternative on cost and applicant-specific factors. The range of alternatives system is deliberately limited to a reasonable number to demonstrate the or to the three most commonly used at operating plants

  19. Sodium Fire Demonstration Facility Design and Operation

    International Nuclear Information System (INIS)

    Cho, Youngil; Kim, Jong-Man; Lee, Jewhan; Hong, Jonggan; Yeom, Sujin; Cho, Chungho; Jung, Min-Hwan; Gam, Da-Young; Jeong, Ji-Young

    2014-01-01

    Although sodium has good characteristics such as high heat transfer rate and stable nuclear property, it is difficult to manage because of high reactivity. Sodium is solid at the room temperature and it easily reacts with oxygen resulting in fire due to the reaction heat. Thus, sodium must be stored in a chemically stable place, i.e., an inert gas-sealed or oil filled vessel. When a sodium fire occurs, the Na 2 O of white fume is formed. It is mainly composed of Na 2 O 2 , NaOH, and Na 2 CO 3 , ranging from 0.1 to several tens of micrometers in size. It is known that the particle size increases by aggregation during floating in air. Thus, the protection method is important and should be considered in the design and operation of a sodium system. In this paper, sodium fire characteristics are described, and the demonstration utility of outbreak of sodium fire and its extinguishing is introduced. In this paper, sodium fire characteristics and a demonstration facility are described. The introduced sodium fire demonstration facility is the only training device used to observe a sodium fire and extinguish it domestically. Furthermore, the type of sodium fire will be diversified with the enhancement of the utility. It is expected that this utility will contribute to experience in the safe treatment of sodium by the handlers

  20. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1979-01-01

    The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

  1. Moderator Demonstration Facility Design and Optimization

    Energy Technology Data Exchange (ETDEWEB)

    McClanahan, Tucker C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gallmeier, Franz X. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Iverson, Erik B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-02-01

    The Spallation Neutron Source (SNS) facility at Oak Ridge National Laboratory (ORNL) is implementing a Moderator Demonstration Facility (MDF) to demonstrate the performance characteristics of advanced moderators central to the Second Target Station (STS) for SNS. The MDF will use the "spare" front-end installation within the SNS accelerator support complex – an ion source, radio-frequency quadrupole (RFQ) accelerator, and medium-energy beam transport (MEBT) chopper - to provide a 2.5 MeV proton beam of peak current 50 mA and maximum pulse length of less than 10 s at a repetition rate of no more than 60 Hz to a suitable neutron-producing target to demonstrate those aspects of moderator performance necessary to meet the goals of the STS design e ort. The accelerator beam parameters are not open to variation beyond that described above - they are fixed by the nature of the spare front-end installation (the Integrated Test Stand Facility; ITSF). Accordingly, there are some neutronic challenges in developing prototypic moderator illumination from a very non-prototypic primary neutron source; the spallation source we are attempting to mimic has an extended neutron source volume approximately 40 cm long (in the direction of the proton beam), approximately 10 cm wide (horizontally transverse to the proton beam) and approximately 5 cm high (vertically transverse to the proton beam), and an isotropic evaporation energy spectrum with mean energy above 1 MeV. In contrast, the primary neutron source available from the 7Li(p,n) reaction (the most prolific at 2.5 MeV proton energy by more than an order of magnitude) is strongly anisotropic, with an energy spectrum that is both strongly dependent on emission angle and kinematically limited to less than 700 keV, and the interaction zone between the incident protons and any target material (neutron-producing or not) is intrinsically limited to a few tens of microns. The MDF will be unique and innovative amongst the world

  2. Generic demonstration plant study (A/E package)

    International Nuclear Information System (INIS)

    Molzen, D.F.

    1979-01-01

    Molzen--Corbin and Associates, Albuquerque, New Mexico, under contract to Sandia Laboratories, has prepared preliminary drawings, descriptive material and a scale model of the demonstration plant. This information will be made available to A/E firms to assist them in the preparation of proposals for complete construction plans and specifications. The four categories for which preliminary work has been prepared consist of structural work, mechanical work, electrical work, and cost estimates. In addition, preliminary specifications, including a written description of the facility consisting of mechanical electrical systems and operations, a description of the safety features, the basic design criteria, three-dimensional sketches, and a scale model of the design have been prepared. The preliminary drawings indicate the required minimum wall thicknesses, overall dimensions and the necessary layout of the removable concrete blocks and slabs required for radiation protection and control

  3. NSS design and plant construction interfaces

    International Nuclear Information System (INIS)

    Stewart, J.J.; Cobb, W.A.

    1976-01-01

    Interface management between NSS design, balance-of-plant design, and plant construction may have a significant effect on schedule stretchout and total plant costs. The paper discusses the importance of the NSS supplier's interface management role, the favorable and unfavorable influencing factors, and examples of interface areas in which experience has demonstrated that problems may arise. Where appropriate, actions are defined to avoid the problems or mitigate the consequences

  4. Demonstration of IGCC features - plant integration and syngas combustion

    Energy Technology Data Exchange (ETDEWEB)

    Hannemann, F.; Huth, M.; Karg, J.; Schiffers, U. [Siemens AG Power Generation (KWU), Erlanger/Muelheim (Germany)

    2000-07-01

    Siemens is involved in three IGCC plants in Europe that are currently in operation. Against the background of the Puertollano and Buggenum plants, some of the specific new features of fully integrated IGCC power plants are discussed, including: requirements and design features of the gas turbine syngas supply system; gas turbine operating experience with air extraction for the air separation unit from the gas turbine air compressor; and design requirements and operational features of the combustion system. 7 refs., 17 figs., 1 tab.

  5. Synthesis gas demonstration plant program, Phase I. Site confirmation report

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    With few reservations, the Baskett, Kentucky site exhibits the necessary characteristics to suggest compatibility with the proposed Synthesis Gas Demonstration Plant Project. An evaluation of a broad range of technical disciplinary criteria in consideration of presently available information indicated generally favorable conditions or, at least, conditions which could be feasibly accommodated in project design. The proximity of the Baskett site to market areas and sources of raw materials as well as a variety of transportation facilities suggests an overall favorable impact on Project economic feasibility. Two aspects of environmental engineering, however, have been identified as areas where the completion or continuation of current studies are required before removing all conditions on site suitability. The first aspect involves the current contradictory status of existing land use and planning ordinances in the site area. Additional investigation of the legality of, and local attitudes toward, these present plans is warranted. Secondly, terrestrial and aquatic surveys of plant and animal life species in the site area must be completed on a seasonal basis to confirm the preliminary conclusion that no exclusionary conditions exist.

  6. Preconceptual design of the gas-phase decontamination demonstration cart

    International Nuclear Information System (INIS)

    Munday, E.B.

    1993-12-01

    Removal of uranium deposits from the interior surfaces of gaseous diffusion equipment will be a major portion of the overall multibillion dollar effort to decontaminate and decommission the gaseous diffusion plants. Long-term low-temperature (LTLT) gas-phase decontamination is being developed at the K-25 Site as an in situ decontamination process that is expected to significantly lower the decontamination costs, reduce worker exposure to radioactive materials, and reduce safeguard concerns. This report documents the preconceptual design of the process equipment that is necessary to conduct a full-scale demonstration of the LTLT method in accordance with the process steps listed above. The process equipment and method proposed in this report are not intended to represent a full-scale production campaign design and operation, since the gas evacuation, gas charging, and off-gas handling systems that would be cost effective in a production campaign are not cost effective for a first-time demonstration. However, the design presented here is expected to be applicable to special decontamination projects beyond the demonstration, which could include the Deposit Recovery Program. The equipment will therefore be sized to a 200 ft size 1 converter (plus a substantial conservative design margin), which is the largest item of interest for gas phase decontamination in the Deposit Recovery Program. The decontamination equipment will allow recovery of the UF 6 , which is generated from the reaction of ClF 3 with the uranium deposits, by use of NaF traps

  7. MHTGR demonstration role in the NRC design certification process

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Jones, G.

    1986-01-01

    A modular high-temperature gas-cooled reactor (MHTGR) design is being developed by the US HTGR Program. Because of the small size of the individual modules that would make up a commercial facility, it appears feasible to design and construct a single-module demonstration plant within the funding constraints on the public and private-sector program participants. Furthermore, the safety margins that can be made inherent to the design permit full-scale testing that could supply a new basis for demonstrating investment protection and safety adequacy to the public, the US Nuclear Regulatory Commission (NRC), and potential users. With this in mind, a Project Definition Study was sponsored by Gas-Cooled Reactor Associates and the Tennessee Valley Authority to study the potential benefits of undertaking such a demonstration project. One of the areas investigated was the potential benefits of such a facility in supporting the NRC design certification process, which is envisioned as a necessary commercialization step for the MHTGR

  8. Evaluation of economical at a uranium enrichment demonstration plant

    International Nuclear Information System (INIS)

    Sugitsue, Noritake

    2001-01-01

    In this report, the economy of technical achievement apply in the uranium enrichment demonstration plant is evaluated. From the evaluation, it can be concluded that the expected purpose was achieved because there was a definite economic prospect to commercial plant. The benefit analysis of thirteen years operation of the uranium enrichment demonstration plant also provides a financial aspect of the uranium enrichment business. Therefore, the performance, price and reliability of the centrifuge is an important factor in the uranium enrichment business. And the continuous development of a centrifuge while considering balance with the development cost is necessary for the business in the future. (author)

  9. Plant aging and design bases documentation

    International Nuclear Information System (INIS)

    Kelly, J.

    1985-01-01

    As interest in plant aging and lifetime extension continues to grow, the need to identify and capture the original design bases for the plant becomes more urgent. Decisions on lifetime extension and availability must be based on a rational understanding of design input, assumptions, and objectives. As operating plant time accumulates, the history of the early design begins to fade. The longer the utility waits, the harder it will be to re-establish the original design bases. Therefore, the time to develop this foundation is now. This paper demonstrates the impact that collecting and maintaining the original design bases of the plant can have on a utility's lifetime extension program. This impact becomes apparent when considering the technical, regulatory and financial aspects of lifetime extension. It is not good enough to know that the design information is buried somewhere in the corporate archives, and that given enough time, it could be retrieved. To be useful to the lifetime extension program, plant design information must be concise, readily available (i.e., retrievable), and easy to use. These objectives can only be met through a systematic program for collecting and presenting plant design documentation. To get the maximum benefit from a lifetime extension program, usable design bases documentation should be available as early in the plant life as possible. It will help identify areas that require monitoring today so that data is available to make rational decisions in the future

  10. Demonstration test for reliability of valves for atomic power plants

    International Nuclear Information System (INIS)

    Hosaka, Shiro

    1978-01-01

    The demonstration test on the reliability of valves for atomic power plants being carried out by the Nuclear Engineering Test Center is reported. This test series is conducted as six-year project from FY 1976 to FY 1981 at the Isogo Test Center. The demonstration test consists of (1) environmental test, (2) reaction force test, (3) vibration test, (4) stress measurement test, (5) operational characteristic test, (6) flow resistance coefficient measuring test, (7) leakage test and (8) safety valve and relief valve test. These contents are explained about the special requirements for nuclear use, for example, the enviornmental condition after the design base accident of PWRs and BWRs, the environmental test sequence for isolation valves of containment vessels under the emergency condition, the seismic test condition for valves of nuclear use, the various stress measurements under thermal transient conditions, the leak test after 500 cycles between the normal operating conditions for PWRs and BWRs and the start up conditions and so on. As for the testing facilities, the whole flow diagram is shown, in which the environmental test section, the vibration test section, the steam test section, the hot water test section, the safety valve test section and main components are included. The specifications of each test section and main components are presented. (Nakai, Y.)

  11. Results from an in-plant demonstration of intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Messick, N.

    1993-01-01

    A learning systems-based reconfigurable controller was demonstrated on the deaerating feedwater heater at the Experimental Breeder Reactor II (EBR-II) on April 1, 1993. Failures of the normal pressure regulating process were introduced by reducing the steam flow to the heater by as much as 10%. The controller maintained pressure in the heater at acceptable levels for several minutes, whereas operator intervention would have otherwise been required within a few seconds. This experiment demonstrates the potential of advanced control techniques for improving safety, reliability, and performance of power plant operations as well as the utility of EBR-II as an experimental power plant controls facility

  12. Conceptual design studies of experimental and demonstration fusion reactors

    International Nuclear Information System (INIS)

    1978-01-01

    Since 1973 the FINTOR Group has been involved in conceptual design studies of TOKAMAK-type fusion reactors to precede the construction of a prototype power reactor plant. FINTOR-1 was the first conceptual design aimed at investigating the main physics and engineering constraints on a minimum-size (both dimensions and thermal power) tokamak experimental reactor. The required plasma energy confinement time as evaluated by various power balance models was compared with the values resulting from different transport models. For the reference design, an energy confinement time ten times smaller than neoclassical was assumed. This also implied a rather high (thermally stable) working temperature (above 20 keV) for the reactor. Other relevant points of the design were: circular plasma cross section, single-null axisymmetric divertor; lithium breeder, stainless steel structures, helium coolant; modular blanket and shield structure; copper-stabilized, superconducting Nb-Ti toroidal field and divertor coils; vertical field and transformer coils inside the toroidal coils; vacuum-tight containment vessel. Solutions involving air and iron transformer cores were compared. These assumptions led to a minimum size reactor with a thermal power of about 100MW and rather large dimensions (major radius of about 9m) similar to those of full-scale power reactors considered in other conceptual studies. The FINTOR-1 analysis was completed by the end of 1976. In 1977 a conceptual design of a Demonstration Power Reactor Plant (FINTOR-D) was started. In this study the main working assumptions differing from those of FINTOR-1 are: non-circular plasma cross section; plasma confinement compatible with trapped ion instabilities; cold (gas) blanket sufficient for wall protection (no divertor); wall loading between 1-3MW/m 2 and thermal power of a few GW. (author)

  13. Early Site Permit Demonstration Program: Plant parameters envelope report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-01

    The Early Site Permit (ESP) Demonstration Program is the nuclear industry`s initiative for piloting the early resolution of siting-related issues before the detailed design proceedings of the combined operating license review. The ESP Demonstration Program consists of three phases. The plant parameters envelopes task is part of Phase 1, which addresses the generic review of applicable federal regulations and develops criteria for safety and environmental assessment of potential sites. The plant parameters envelopes identify parameters that characterize the interface between an ALWR design and a potential site, and quantify the interface through values selected from the Utility Requirements Documents, vendor design information, or engineering assessments. When augmented with site-specific information, the plant parameters envelopes provide sufficient information to allow ESPs to be granted based on individual ALWR design information or enveloping design information for the evolutionary, passive, or generic ALWR plants. This document is expected to become a living document when used by future applicants.

  14. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  15. Waste water pilot plant research, development, and demonstration permit application

    International Nuclear Information System (INIS)

    1993-03-01

    This permit application has been prepared to obtain a research, development, and demonstration permit to perform pilot-scale treatability testing on the 242-A Evaporator process condensate waste water effluent stream. It provides the management framework, and controls all the testing conducted in the waste water pilot plant using dangerous waste. It also provides a waste acceptance envelope (upper limits for selected constituents) and details the safety and environmental protection requirements for waste water pilot plant testing. This permit application describes the overall approach to testing and the various components or requirements that are common to all tests. This permit application has been prepared at a sufficient level of detail to establish permit conditions for all waste water pilot plant tests to be conducted

  16. Reactive power control with CHP plants - A demonstration

    DEFF Research Database (Denmark)

    Nyeng, Preben; Østergaard, Jacob; Andersen, Claus A.

    2010-01-01

    power rating of 7.3 MW on two synchronous generators. A closed-loop control is implemented, that remote controls the CHP plant to achieve a certain reactive power flow in a near-by substation. The solution communicates with the grid operator’s existing SCADA system to obtain measurements from...... lines to underground cables has changed the reactive power balance, and third, the TSO has introduced restrictions in the allowed exchange of reactive power between the transmission system and distribution grids (known as the Mvar-arrangement). The demonstration includes a CHP plant with an electric......In this project the potential for ancillary services provision by distributed energy resources is investigated. Specifically, the provision of reactive power control by combined heat and power plants is examined, and the application of the new standard for DER communication systems, IEC 61850...

  17. Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

    2011-01-01

    JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

  18. Demonstration of risk-based approaches to nuclear plant regulation

    International Nuclear Information System (INIS)

    Rahn, F.J.; Sursock, J.P.; Darling, S.S.; Oddo, J.M.

    1993-01-01

    This paper describes generic technical support EPRI is providing to the nuclear power industry relative to its recent initiatives in the area of risk-based regulations (RBR). A risk-based regulatory approach uses probabilistic risk assessment (PRA), or similar techniques, to allocate safety resources commensurate with the risk posed by nuclear plant operations. This approach will reduce O ampersand M costs, and also improve nuclear plant safety. In order to enhance industry, Nuclear Regulatory Commission (NRC) and public confidence in RBR, three things need to be shown: (1) manpower/resource savings are significant for both NRC and industry; (2) the process is doable in a reasonable amount of time; and (3) the process, if uniformly applied, results in demonstrably cheaper power and safer plants. In 1992, EPRI performed a qualitative study of the key RBR issues contributing to high O ampersand M costs. The results are given on Table 1. This study is being followed up by an in-depth quantitative cost/benefit study to focus technical work on producing guidelines/procedures for licensing submittals to NRC. The guidelines/procedures necessarily will be developed from successful demonstration projects such as the Fitzpatrick pilot plant study proposed by the New York Power Authority and other generic applications. This paper presents three examples: two motor operated valve projects performed by QUADREX Energy Services Corporation working with utilities in responding to NRC Generic Letter 89-10, and a third project working with Yankee Atomic Electric Company on service water systems at a plant in its service system. These demonstration projects aim to show the following: (1) the relative ease of putting together a technical case based on RBR concepts; (2) clarity in differentiating the various risk trade-offs, and in communicating overall reductions in risk with NRC; and (3) improved prioritization of NRC directives

  19. Plant design and beam utilization

    International Nuclear Information System (INIS)

    Svendsen, E.B.

    1983-01-01

    Plant design and beam utilization are two things closely tied together: without a proper plant design, one can never get good beam utilization. When a company decides to build an irradiation facility, there are some major decisions to be made right in the beginning. These decisions can be most important for the long-term success or failure of the irradiation facility, because the company normally will have to live with these decisions during the whole life-time of the irradiation equipment. To start with the decision has to be made whether to select a cobalt-60 irradiation plant or an accelerator irradiation plant. This decision can only be reached after a careful study of the products and the 'weight' and the material of the products the company wants to irradiate. As an old accelerator-man, I tend to personally favor accelerators, although I am very impressed by the newer cobalt-60 pallet irradiation plants from A.E.C.L. I believe that they have a great future in the emerging field of food irradiation. As I have primarily been involved with accelerators during the last 14 years, this paper is only dealing with different design approaches and utilizations of accelerator-plants. (author)

  20. Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

    International Nuclear Information System (INIS)

    Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

    2011-01-01

    This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

  1. Implementing plant design changes economically

    International Nuclear Information System (INIS)

    Seamans, L.D.

    1994-01-01

    Consumers Power Company, as is the case with most utilities, is challenged by the necessity to control non-modification design changes. At the Consumers' Palisades Nuclear Plant, the challenge was met by the establishment of a project team that developed an innovative comprehensive design control process called Functional Equivalent Substitution (FES). This paper describes: (1) the project team composition; (2) the process development methodology; and (3) the implementation results

  2. The Evritania (Greece) demonstration plant of biomass pyrolysis

    Energy Technology Data Exchange (ETDEWEB)

    Zabaniotou, A.A.; Karabela, A.J. [Aristotle University of Thessaloniki (Greece). Dept. of Chemical Engineering and Chemical Process Engineering Research Institute

    1999-06-01

    This paper is focused on describing the Evritania demonstration plant for pyrolysis of forestry biomass. This plant was constructed in the village of Voulpi, district of Evritania, in central Greece, in 1995, with a threefold purpose; development of know-how, forest fire prevention and rural development. The products are charcoal and bio-oil. The plant capacity is 1200-1450 kg/h of wet biomass and the pyrolysis temperature is approx. 400 deg C. The raw material used is Arbutus unedo, which is an evergreen broad-leaf tree which covers the area. Other agricultural waste could also be used, such as olive pits and cuttings, almond shells and cotton kernels. The paper includes the conceptual process flow sheet, the changes and improvements made during the trial phase, data from the start-up phase, and product characteristics. Comparison of the process with the Alten process is presented. Additionally, comparisons are made of product characteristics with those from other pyrolysis processes. In general, the results obtained are encouraging even though several improvements of the pilot plant are required. (author)

  3. Vitrification process equipment design for the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Chapman, C.C.; Drosjack, W.P.

    1988-10-01

    The vitrification process and equipment design is nearing completion for the West Valley Project. This report provides the basis and current status for the design of the major vessels and equipment within the West Valley Vitrification Plant. A review of the function and key design features of the equipment is also provided. The major subsystems described include the feed preparation and delivery systems, the melter, the canister handling systems, and the process off-gas system. 11 refs., 33 figs., 4 tabs

  4. Coupling solar photo-Fenton and biotreatment at industrial scale: Main results of a demonstration plant

    International Nuclear Information System (INIS)

    Malato, Sixto; Blanco, Julian; Maldonado, Manuel I.; Oller, Isabel; Gernjak, Wolfgang; Perez-Estrada, Leonidas

    2007-01-01

    This paper reports on the combined solar photo-Fenton/biological treatment of an industrial effluent (initial total organic carbon, TOC, around 500 mg L -1 ) containing a non-biodegradable organic substance (α-methylphenylglycine at 500 mg L -1 ), focusing on pilot plant tests performed for design of an industrial plant, the design itself and the plant layout. Pilot plant tests have demonstrated that biodegradability enhancement is closely related to disappearance of the parent compound, for which a certain illumination time and hydrogen peroxide consumption are required, working at pH 2.8 and adding Fe 2+ = 20 mg L -1 . Based on pilot plant results, an industrial plant with 100 m 2 of CPC collectors for a 250 L/h treatment capacity has been designed. The solar system discharges the wastewater (WW) pre-treated by photo-Fenton into a biotreatment based on an immobilized biomass reactor. First, results of the industrial plant are also presented, demonstrating that it is able to treat up to 500 L h -1 at an average solar ultraviolet radiation of 22.9 W m -2 , under the same conditions (pH, hydrogen peroxide consumption) tested in the pilot plant

  5. Startup, testing, and operation of the Santa Clara 2MW direct carbonate fuel cell demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Skok, A.J.; Leo, A.J. [Fuel Cell Engineering Corp., Danbury, CT (United States); O`Shea, T.P. [Santa Clara Demonstration Project, CA (United States)

    1996-12-31

    The Santa Clara Demonstration Project (SCDP) is a collaboration between several utility organizations, Fuel Cell Engineering Corporation (FCE), and the U.S. Dept. Of Energy aimed at the demonstration of Energy Research Corporation`s (ERC) direct carbonate fuel cell (DFC) technology. ERC has been pursuing the development of the DFC for commercialization near the end of this decade, and this project is an integral part of the ERC commercialization effort. The objective of the Santa Clara Demonstration Project is to provide the first full, commercial scale demonstration of this technology. The approach ERC has taken in the commercialization of the DFC is described in detail elsewhere. An aggressive core technology development program is in place which is focused by ongoing interaction with customers and vendors to optimize the design of the commercial power plant. ERC has selected a 2.85 MW power plant unit for initial market entry. Two ERC subsidiaries are supporting the commercialization effort: the Fuel Cell Manufacturing Corporation (FCMC) and the Fuel Cell Engineering Corporation (FCE). FCMC manufactures carbonate stacks and multi-stack modules, currently from its production facility in Torrington, CT. FCE is responsible for power plant design, integration of all subsystems, sales/marketing, and client services. FCE is serving as the prime contractor for the design, construction, and testing of the SCDP Plant. FCMC has manufactured the multi-stack submodules used in the DC power section of the plant. Fluor Daniel Inc. (FDI) served as the architect-engineer subcontractor for the design and construction of the plant and provided support to the design of the multi-stack submodules. FDI is also assisting the ERC companies in commercial power plant design.

  6. Savannah River Plant low-level waste incinerator demonstration

    International Nuclear Information System (INIS)

    Tallman, J.A.

    1984-01-01

    A two-year demonstration facility was constructed at the Savannah River Plant (SRP) to incinerate suspect contaminated solid and low-level solvent wastes. Since startup in January 1984, 4460 kilograms and 5300 liters of simulated (uncontaminated) solid and solvent waste have been incinerated to establish the technical and operating data base for the facility. Combustion safeguards have been enhanced, process controls and interlocks refined, some materials handling problems identified and operating experience gained as a result of the 6 month cold run-in. Volume reductions of 20:1 for solid and 25:1 for solvent waste have been demonstrated. Stack emissions (NO 2 , SO 2 , CO, and particulates) were only 0.5% of the South Carolina ambient air quality standards. Radioactive waste processing is scheduled to begin in July 1984. 2 figures, 2 tables

  7. Field Demonstration of Fuel Crud Filtration System at Ulchin Plant

    International Nuclear Information System (INIS)

    Kang, Duk-Won; Lee, Doo-Ho; Park, Jong-Youl; Choi, In-Kyu

    2007-01-01

    Crud deposited onto the fuel assemblies in nuclear power plants was not a serious problem until an upper core flux depression named Axial Offset Anomaly (AOA) was found at Callaway, USA in 1989. Though the mechanism of an AOA is not completely understood, crud is believed to be a key component of initiating AOA. After the sufficient amount of corrosion products in the reactor cooling system are deposited on the fuel clad by a sub-cooled nucleate boiling, boron is adsorbed in the crud. Thus a measurable reduction in the neutron flux occurs which causes an AOA problem. A filtration system has been developed to remove the fuel crud from irradiated fuel assemblies for mitigating the axial offset anomaly under a technical cooperation agreement with DEI (Dominion Engineering Inc.). This filtration system with a fuel cleaning fixture was successfully demonstrated at Ulchin plant unit 2. Within several minutes, detachable crud deposits were effectively removed from the clad surfaces of the fuel assembly. Also, to characterize the crud particles for each fuel assembly, a small crud sampling device and radiation monitor devices were connected to the filtration system during the cleaning operation. In this study, we completed a functional test and demonstration of an ultrasonic fuel cleaning system by using four spent fuel assemblies. It took only 5 minutes to remove the fuel crud from each fuel assembly. In addition, collective dose rates indicated an average of 8 R/Hr per assembly

  8. 1000kW on-site PAFC power plant development and demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Satomi, Tomohide; Koike, Shunichi [Phosphoric Acid Fuel Cell Technology Research Association (PAFC-TRA), Osaka (Japan); Ishikawa, Ryou [New Energy and Industrial Technology Development Organization (NEDO), Tokyo (Japan)

    1996-12-31

    Phosphoric Acid Fuel Cell Technology Research Association (PAFC-TRA) and New Energy and Industrial Technology Development Organization (NEDO) have been conducting a joint project on development of a 5000kW urban energy center type PAFC power plant (pressurized) and a 1000kW on-site PAFC power plant (non-pressurized). The objective of the technical development of 1000kW on-site PAFC power plant is to realize a medium size power plant with an overall efficiency of over 70% and an electrical efficiency of over 36%, that could be installed in a large building as a cogeneration system. The components and system integration development work and the plant design were performed in 1991 and 1992. Manufacturing of the plant and installation at the test site were completed in 1994. PAC test was carried out in 1994, and generation test was started in January 1995. Demonstration test is scheduled for 1995 and 1996.

  9. Demonstration tokamak fusion power plant for early realization of net electric power generation

    International Nuclear Information System (INIS)

    Hiwatari, R.; Okano, K.; Asaoka, Y.; Shinya, K.; Ogawa, Y.

    2005-01-01

    A demonstration tokamak fusion power plant Demo-CREST is proposed as the device for early realization of net electric power generation by fusion energy. The plasma configuration for Demo-CREST is optimized to satisfy the electric breakeven condition (the condition for net electric power, P e net = 0 MW) with the plasma performance of the ITER reference operation mode. This optimization method is considered to be suitable for the design of a demonstration power plant for early realization of net electric power generation, because the demonstration power plant has to ensure the net electric generation. Plasma performance should also be more reliably achieved than in past design studies. For the plasma performance planned in the present ITER programme, net electric power from 0 to 500 MW is possible with Demo-CREST under the following engineering conditions: maximum magnetic field 16 T, thermal efficiency 30%, NBI system efficiency 50% and NBI current drive power restricted to 200 MW. By replacing the blanket system with one of higher thermal efficiency, a net electric power of about 1000 MW is also possible so that the performance of the commercial plant with Demo-CREST can also be studied from the economic point of view. The development path from the experimental reactor 'ITER' to the commercial plant 'CREST' through the demonstration power plant 'Demo-CREST' is proposed as an example of the fast track concept. (author)

  10. Overall plant design of PWRs

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1980-01-01

    In the present lecture the main components and safety related systems are described, to get a general overview about the safety measures in a PWR. The idea to introduce safety systems is to protect the nuclear reactor core against the so-called design accidents and to prevent the release of activity to the environment. Furthermore the operation personnel has to be protected against radioactive contamination. All redundant and diversified safety measures used in a nuclear power station ensure reliable and safe operation of the plant in all modes of operation. To minimize the operational risk to an extended minimum besides active safety systems a lot of passive safety barriers are foreseen. With the design and construction, tests and quality assurance measures are performed to assure a safe plant operation. (orig.)

  11. Design of nuclear power plants

    International Nuclear Information System (INIS)

    Lobo, C.G.

    1987-01-01

    The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.) [pt

  12. Designing nuclear power plants for improved operation and maintenance

    International Nuclear Information System (INIS)

    1996-09-01

    The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig

  13. Designing nuclear power plants for improved operation and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig.

  14. Safety demonstration test on solvent fire in fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nishio, Gunji; Hashimoto, Kazuichiro

    1989-03-01

    This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

  15. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  16. Evaluation of the uranium enrichment demonstration plant project

    International Nuclear Information System (INIS)

    Sugitsue, Noritake

    2001-01-01

    In this report, the organization system of the uranium enrichment business is evaluated, based on the operation of the uranium enrichment demonstration plant. As a result, in uranium enrichment technology development or business, it was acknowledged that maintenance of the organization which has the Trinity of a research/engineering/operation was necessary in an industrialization stage by exceptional R and D cycle. Japan Nuclear Fuel Ltd. (JNFL) set up the Rokkashomura Aomori Uranium Enrichment Research and Development Center in November 2000. As a result, the system that company directly engaged in engineering development was prepared. And results obtained in this place is expected toward certain establishment of the uranium enrichment business of Japan. (author)

  17. Industrial Fuel Gas Demonstration Plant Program: environmental permit compliance plan

    Energy Technology Data Exchange (ETDEWEB)

    Bodamer, Jr., James W.; Bocchino, Robert M.

    1979-11-01

    This Environmental Permit Compliance Plan is intended to assist the Memphis Light, Gas and Water Division in acquiring the necessary environmental permits for their proposed Industrial Fuel Gas Demonstration Plant in a time frame consistent with the construction schedule. Permits included are those required for installation and/or operation of gaseous, liquid and solid waste sources and disposal areas. Only those permits presently established by final regulations are described. The compliance plan describes procedures for obtaining each permit from identified federal, state and local agencies. The information needed for the permit application is presented, and the stepwise procedure to follow when filing the permit application is described. Information given in this plan was obtained by reviewing applicable laws and regulations and from telephone conversations with agency personnel on the federal, state and local levels. This Plan also presents a recommended schedule for beginning the work necessary to obtain the required environmental permits in order to begin dredging operations in October, 1980 and construction of the plant in September, 1981. Activity for several key permits should begin as soon as possible.

  18. Power-up of Fugen reactor and development of demonstration plant

    International Nuclear Information System (INIS)

    Sawai, Sadamu; Akebi, Michio; Yazaki, Akira.

    1979-06-01

    The Fugen Nuclear Power Station is the 165 MWe prototype plant characterized by heavy water-moderated, boiling light water-cooled, pressure tube type, and was developed by the Power Reactor and Nuclear Fuel Development Corporation, Japan. The plant went into commercial operation on March 20, 1979, in Tsuruga, Fukui Prefecture. Some delay in the overall schedule occurred due to the shortage of cement caused by the oil crisis, more stringent regulations as the result of stress corrosion cracking experienced in BWRs, and design modifications. All functional tests, the final check-up of the whole plant, and remaining modifying works had been completed by March 10, 1978. The minimum criticality was achieved with 22 mixed oxide fuel assemblies on March 20, 1978. Thereafter, the tests on reactor physics, plant dynamics, the performances of components and systems, and radiation and water chemistry have been carried out. 5 MWe was sent to grid system for the first time on July 29, 1978. The commercial operation of the plant was licenced by the Government on March 30, 1979. The conceptual design of the 600 MWe demonstration plant was finished in early 1979, and the detailed design is to be carried out in 1979 and 1980. The main design principle was incorporated in the conceptual design, but some modifications are to be made to reduce the power cost and to facilitate the easy maintenance. (Kako, I.)

  19. The Continuous Wave Deuterium Demonstrator (CWDD) design and status

    Energy Technology Data Exchange (ETDEWEB)

    Todd, A.M.M. (Grumman Space and Electronics Corp., Princeton, NJ (United States)); Nightingale, M.P.S. (AEA Industrial Technology, Culham (United Kingdom)); Yule, T.J. (Argonne National Lab., IL (United States))

    1992-01-01

    The design of the Continuous Wave Deuterium Demonstrator (CWDD) and the status of the fabricated hardware is presented. The CWDD is a high brightness, 352 MHz, CW linear accelerator designed to deliver a 7.54 MeV, 80 mA D[sup [minus

  20. Experience gained with the Synroc demonstration plant at ANSTO and its relevance to plutonium immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Jostsons, A.; Ridal, A.; Mercer, D.J.; Vance, E.R.L. [Australian Nuclear Science and Technology Organisation, Menai (Australia)

    1996-05-01

    The Synroc Demonstration Plant (SDP) was designed and constructed at Lucas Heights to demonstrate the feasibility of Synroc production on a commercial scale (10 kg/hr) with simulated Purex liquid HLW. Since commissioning of the SDP in 1987, over 6000 kg of Synroc has been fabricated with a range of feeds and waste loadings. The SDP utilises uniaxial hot-pressing to consolidate Synroc. Pressureless sintering and hot-isostatic pressing have also been studied at smaller scales. The results of this extensive process development have been incorporated in a conceptual design for a radioactive plant to condition HLW from a reprocessing plant with a capacity to treat 800 tpa of spent LWR fuel. Synroic containing TRU, including Pu, and fission products has been fabricated and characterised in a glove-box facility and hot cells, respectively. The extensive experience in processing of Synroc over the past 15 years is summarised and its relevance to immobilization of surplus plutonium is discussed.

  1. Advanced plant design recommendations from Cook Nuclear Plant experience

    International Nuclear Information System (INIS)

    Zimmerman, W.L.

    1993-01-01

    A project in the American Electric Power Service Corporation to review operating and maintenance experience at Cook Nuclear Plant to identify recommendations for advanced nuclear plant design is described. Recommendations so gathered in the areas of plant fluid systems, instrument and control, testing and surveillance provisions, plant layout of equipment, provisions to enhance effective maintenance, ventilation systems, radiological protection, and construction, are presented accordingly. An example for a design review checklist for effective plant operations and maintenance is suggested

  2. Industrial Fuel Gas Demonstration Plant Program. Task III, Demonstration plant safety, industrial hygiene, and major disaster plan (Deliverable No. 35)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-03-01

    This Health and Safety Plan has been adopted by the IFG Demonstration Plant managed by Memphis Light, Gas and Water at Memphis, Tennessee. The plan encompasses the following areas of concern: Safety Plan Administration, Industrial Health, Industrial Safety, First Aid, Fire Protection (including fire prevention and control), and Control of Safety Related Losses. The primary objective of this plan is to achieve adequate control of all potentially hazardous activities to assure the health and safety of all employees and eliminate lost work time to both the employees and the company. The second objective is to achieve compliance with all Federal, state and local laws, regulations and codes. Some thirty specific safe practice instruction items are included.

  3. The design of Chp plants

    International Nuclear Information System (INIS)

    Tomassetti, G.

    2001-01-01

    Chp is considered with a bottom-up view, as the most efficient way to satisfy the needs of the users. In order to achieve optimal results a particular care must be used in analyzing the thermal and electrical loads and their interactions. On this basis and taking into account the relationships among the user and the suppliers of electricity, fuels and heat, the energy market structure, the cost of energy and the tax assessment it is possible to properly design Chp plants with benefits for the users [it

  4. Conceptual design of the JAERI demonstration fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Tone, T.; Seki, Y.

    1976-01-01

    Conceptual design of a tokamak demonstration fusion reactor is carried out. This design is an extended and improved version of the previous design which was presented at the 5th IAEA Conference. The main design parameters are as follows: the reactor thermal power 2000 MW, torus radius 10.5 m, plasma radius 2.7 m, first wall radius 3.0 m, toroidal magnetic field on axis 6T, blanket fertile material Li 2 O, coolant He, structural material Mo-alloy and tritium breeding ratio 1.2

  5. PA activity by using nuclear power plant safety demonstration and analysis

    International Nuclear Information System (INIS)

    Tsuchiya, Mitsuo; Kamimae, Rie

    1999-01-01

    INS/NUPEC presents one of Public acceptance (PA) methods for nuclear power in Japan, 'PA activity by using Nuclear Power Plant Safety Demonstration and Analysis', by using one of videos which is explained and analyzed accident events (Loss of Coolant Accident). Safety regulations of The National Government are strictly implemented in licensing at each of basic design and detailed design. To support safety regulation activities conducted by the National Government, INS/NLTPEC continuously implement Safety demonstration and analysis. With safety demonstration and analysis, made by assuming some abnormal conditions, what impacts could be produced by the assumed conditions are forecast based on specific design data on a given nuclear power plants. When analysis results compared with relevant decision criteria, the safety of nuclear power plants is confirmed. The decision criteria are designed to help judge if or not safety design of nuclear power plants is properly made. The decision criteria are set in the safety examination guidelines by taking sufficient safety allowance based on the latest technical knowledge obtained from a wide range of tests and safety studies. Safety demonstration and analysis is made by taking the procedure which are summarized in this presentation. In Japan, various PA (Public Acceptance) pamphlets and videos on nuclear energy have been published. But many of them focused on such topics as necessity or importance of nuclear energy, basic principles of nuclear power generation, etc., and a few described safety evaluation particularly of abnormal and accident events in accordance with the regulatory requirements. In this background, INS/NUPEC has been making efforts to prepare PA pamphlets and videos to explain the safety of nuclear power plants, to be simple and concrete enough, using various analytical computations for abnormal and accident events. In results, PA activity of INS/NUPEC is evaluated highly by the people

  6. Conceptual design of a demonstration reactor for electric power generation

    International Nuclear Information System (INIS)

    Asaoka, Y.; Hiwatari, R.; Okano, K.; Ogawa, Y.; Ise, H.; Nomoto, Y.; Kuroda, T.; Mori, S.; Shinya, K.

    2005-01-01

    Conceptual study on a demonstration plant for electric power generation, named Demo-CREST, was conducted based on the consideration that a demo-plant should have capacities both (1) to demonstrate electric power generation in a plant scale with moderate plasma performance, which will be achieved in the early stage of the ITER operation, and foreseeable technologies and materials and (2) to have a possibility to show an economical competitiveness with advanced plasma performance and high performance blanket systems. The plasma core was optimized to be a minimum size for both net electric power generation with the ITER basic plasma parameters and commercial-scale generation with advance plasma parameters, which would be attained by the end of ITER operation. The engineering concept, especially the breeding blanket structure and its maintenance scheme, is also optimized to demonstrate the tritium self-sustainability and maintainability of in-vessel components. Within the plasma performance as planned in the present ITER program, the net electric power from 0 MW to 500 MW is possible with the basic blanket system under the engineering conditions of maximum magnetic field 16 T, NBI system efficiency 50%, and NBI current drive power restricted to 200 MW. Capacities of stabilization of reversed shear plasma and the high thermal efficiency are additional factors for optimization of the advanced blanket. By replacing the blanket system with the advanced one of higher thermal efficiency, the net electric power of about 1000 MW is also possible so that the economic performance toward the commercial plant can be also examined with Demo-CREST. (author)

  7. Demonstration plant of smoke treatment by electron beam irradiation

    International Nuclear Information System (INIS)

    Kawamura, Keita

    1989-01-01

    The acid rain caused by sulfur oxides and nitrogen oxides has become the large social problem as it damages forests, lakes and agricultural crops and also buildings in Europe and America. In such circumstances, concern has been expressed in various countries on the smoke treatment technology, EBA process, which removes the sulfur oxides and nitrogen oxides contained in smoke simultaneously by irradiating electron beam on the smoke which is exhausted from power station boilers and industrial boilers and mainly causes acid rain. The research and development of this technology were begun in 1971 based on the original idea of Ebara Corp., and from 1972, those were advanced as the joint research with Japan Atomic Energy Research Institute. Thereafter, by the joint research with the technical research association on prevention of nitrogen oxides in iron and steel industry, by ammonia addition and irradiation process, the desulfurization and denitration performance was heightened, and the byproduct was successfully captured as powder, in this way, the continuous dry treatment process was established. The demonstration test plant was constructed in a coal-firing power station in Indiana, USA, and the trial operation was carried out from 1985 for two years. (K.I.)

  8. Possible futures for the development of a fusion demonstration plant

    International Nuclear Information System (INIS)

    Nichols, S.P.

    1976-01-01

    As indicated by the Fusion Planning Bulletins, the Division of Controlled Thermonuclear Research is becoming involved in planning with alternative scenarios. The Center for Energy Studies at the University of Texas at Austin has been involved with such planning for several years and has examined various scenarios for fusion power development using the Partitive Analytical Forecasting (PAF) technique. The most recent studies compare the long-term plan presented in WASH-1290, Fusion Power by Magnetic Confinement, with other plans that have been proposed, such as the plan proposed by Kulcinski and Conn of the University of Wisconsin. The study indicates that some of the alternative plans do have possibilities to shorten the required time for the completion of a demonstration fusion plant without increased costs or a decrease in the likelihood of success. The current efforts of the project are in the planning exercises recently completed by committees set up by the DCTR. Further comparisons of alternative scenarios will be performed as part of this effort

  9. Design alternatives report for the cesium removal demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Walker, J.F. Jr.; Youngblood, E.L.

    1995-09-01

    The Cesium Removal Demonstration (CRD) project will use liquid low-level waste (LLLW) stored in the Oak Ridge National Laboratory Melton Valley Storage Tanks to demonstrate cesium removal from sodium nitrate-based supernates. This report presents the results of a conceptual design study to scope the alternatives for conducting the demonstration at ORNL. Factors considered included (1) sorbent alternatives, (2) facility alternatives, (3) process alternatives, (4) process disposal alternatives, and (5) relative cost comparisons. Recommendations included (1) that design of the CRD system move forward based on information obtained to date from tests with Savannah River Resin, (2) that the CRD system be designed so it could use crystalline silicotitanates (CST) if an engineered form of CST becomes available prior to the CRD, (3) that the system be designed without the capability for resin regeneration, (4) that the LLLW solidification facility be used for the demonstration (5) that vitrification of the loaded resins from the CRD be demonstrated at the Savannah River Site, and (6) that permanent disposal of the loaded and/or vitrified resin at the Nevada Test Site be pursued.

  10. Design alternatives report for the cesium removal demonstration

    International Nuclear Information System (INIS)

    Walker, J.F. Jr.; Youngblood, E.L.

    1995-09-01

    The Cesium Removal Demonstration (CRD) project will use liquid low-level waste (LLLW) stored in the Oak Ridge National Laboratory Melton Valley Storage Tanks to demonstrate cesium removal from sodium nitrate-based supernates. This report presents the results of a conceptual design study to scope the alternatives for conducting the demonstration at ORNL. Factors considered included (1) sorbent alternatives, (2) facility alternatives, (3) process alternatives, (4) process disposal alternatives, and (5) relative cost comparisons. Recommendations included (1) that design of the CRD system move forward based on information obtained to date from tests with Savannah River Resin, (2) that the CRD system be designed so it could use crystalline silicotitanates (CST) if an engineered form of CST becomes available prior to the CRD, (3) that the system be designed without the capability for resin regeneration, (4) that the LLLW solidification facility be used for the demonstration (5) that vitrification of the loaded resins from the CRD be demonstrated at the Savannah River Site, and (6) that permanent disposal of the loaded and/or vitrified resin at the Nevada Test Site be pursued

  11. ELISA, a demonstrator environment for information systems architecture design

    Science.gov (United States)

    Panem, Chantal

    1994-01-01

    This paper describes an approach of reusability of software engineering technology in the area of ground space system design. System engineers have lots of needs similar to software developers: sharing of a common data base, capitalization of knowledge, definition of a common design process, communication between different technical domains. Moreover system designers need to simulate dynamically their system as early as possible. Software development environments, methods and tools now become operational and widely used. Their architecture is based on a unique object base, a set of common management services and they host a family of tools for each life cycle activity. In late '92, CNES decided to develop a demonstrative software environment supporting some system activities. The design of ground space data processing systems was chosen as the application domain. ELISA (Integrated Software Environment for Architectures Specification) was specified as a 'demonstrator', i.e. a sufficient basis for demonstrations, evaluation and future operational enhancements. A process with three phases was implemented: system requirements definition, design of system architectures models, and selection of physical architectures. Each phase is composed of several activities that can be performed in parallel, with the provision of Commercial Off the Shelves Tools. ELISA has been delivered to CNES in January 94, currently used for demonstrations and evaluations on real projects (e.g. SPOT4 Satellite Control Center). It is on the way of new evolutions.

  12. Equipment designs for the spent LWR fuel dry storage demonstration

    International Nuclear Information System (INIS)

    Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

    1980-01-01

    In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations

  13. Current status and technical description of Chinese 2 x 250 MWth HTR-PM demonstration plant

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Wu Zongxin; Wang Dazhong; Xu Yuanhui; Sun Yuliang; Li Fu; Dong Yujie

    2009-01-01

    After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with much improved safety features. Following the successful construction and a most gratifying operation of the 10 MW th high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed and designed an HTR demonstration plant, called the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out with industry partners from China and in collaboration of experts worldwide, closely follows the design principles of the HTR-10. Due to intensive engineering and R and D efforts since 2001, the basic design of the HTR-PM has been finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR) has been compiled. The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one comprising of a single zone 250 MW th pebble-bed modular reactor and a steam generator. The two NSSS modules feed one steam turbine and generate an electric power of 210 MW. A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is closely based on the technology of the HTR-10 fuel production line. The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident management procedures and will not require any need for offsite emergency measures. According to the current schedule of the project the completion date of the demonstration plant will be around 2013. The reactor site has been evaluated and approved; the procurement of long-lead components has already been started. After the successful operation of the demonstration plant

  14. Westinghouse AP600 advanced nuclear plant design

    International Nuclear Information System (INIS)

    Gangloff, W.

    1999-01-01

    As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

  15. DESIGN OF SMALL AUTOMATION WORK CELL SYSTEM DEMONSTRATIONS

    International Nuclear Information System (INIS)

    TURNER, C.; PEHL, J.

    2000-01-01

    The introduction of automation systems into many of the facilities dealing with the production, use and disposition of nuclear materials has been an ongoing objective. Many previous attempts have been made, using a variety of monolithic and, in some cases, modular technologies. Many of these attempts were less than successful, owing to the difficulty of the problem, the lack of maturity of the technology, and over optimism about the capabilities of a particular system. Consequently, it is not surprising that suggestions that automation can reduce worker Occupational Radiation Exposure (ORE) levels are often met with skepticism and caution. The development of effective demonstrations of these technologies is of vital importance if automation is to become an acceptable option for nuclear material processing environments. The University of Texas Robotics Research Group (UTRRG) has been pursuing the development of technologies to support modular small automation systems (each of less than 5 degrees-of-freedom) and the design of those systems for more than two decades. Properly designed and implemented, these technologies have a potential to reduce the worker ORE associated with work in nuclear materials processing facilities. Successful development of systems for these applications requires the development of technologies that meet the requirements of the applications. These application requirements form a general set of rules that applicable technologies and approaches need to adhere to, but in and of themselves are generally insufficient for the design of a specific automation system. For the design of an appropriate system, the associated task specifications and relationships need to be defined. These task specifications also provide a means by which appropriate technology demonstrations can be defined. Based on the requirements and specifications of the operations of the Advanced Recovery and Integrated Extraction System (ARIES) pilot line at Los Alamos National

  16. Reality testing a plant design 'virtually' anywhere

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    The development of a new world-wide-web compatible information system known as HyperPlant will allow users to navigate real-time three-dimensional plant design and contraction software. It is anticipated that corporate Intranets will be created to facilitate computer-aided design of industrial plants such as piping routes, process schematics, fabrication drawings, and allow use of PDMS (the Plant Design Management System). HyperPlant can also assist in plant commissioning and operation as well as for planning operation and maintenance procedures. (UK)

  17. Reliability analysis techniques in power plant design

    International Nuclear Information System (INIS)

    Chang, N.E.

    1981-01-01

    An overview of reliability analysis techniques is presented as applied to power plant design. The key terms, power plant performance, reliability, availability and maintainability are defined. Reliability modeling, methods of analysis and component reliability data are briefly reviewed. Application of reliability analysis techniques from a design engineering approach to improving power plant productivity is discussed. (author)

  18. Shielding design for better plant availability

    International Nuclear Information System (INIS)

    Biro, G.G.

    1975-01-01

    Design methods are described for providing a shield system for nuclear power plants that will facilitate maintenance and inspection, increase overall plant availability, and ensure that man-rem exposures are as low as practicable

  19. 40 CFR 63.1585 - How does an industrial POTW treatment plant demonstrate compliance?

    Science.gov (United States)

    2010-07-01

    ... Works Industrial Potw Treatment Plant Description and Requirements § 63.1585 How does an industrial POTW treatment plant demonstrate compliance? (a) An existing industrial POTW treatment plant demonstrates... §§ 63.1586 through 63.1590. Non-industrial POTW Treatment Plant Requirements ...

  20. Design study on the helium engineering demonstration loop (HENDEL)

    International Nuclear Information System (INIS)

    Aochi, Tetsuo; Yasuno, Takehiko; Muto, Yasushi; Suzuki, Kunihiko

    1977-11-01

    Four reference studies made on Helium Engineering Demonstration Loop (HENDEL) are described. HENDEL is used in confirmation of the designs of VHTR components such as reactor structure, core structure, intermediate heat exchanger and piping. It consists of mother loop, adapter section and four test sections for fuel stack, reactor support and insulation structure, core structure and high temperature heat transfer component respectively. System and component designs of the mother and adapter section and preliminary designs of the four test sections are shown. And, the plans of operation, instrumentation, control, safety, utilities (electricity, cooling water and helium gas) and construction schedule of HENDEL and research and development of the test sections are also briefed. (auth.)

  1. Design, integration and demonstration of a 50 W JP8/kerosene fueled portable SOFC power generator

    Science.gov (United States)

    Cheekatamarla, Praveen K.; Finnerty, Caine M.; Robinson, Charles R.; Andrews, Stanley M.; Brodie, Jonathan A.; Lu, Y.; DeWald, Paul G.

    A man-portable solid oxide fuel cell (SOFC) system integrated with desulfurized JP8 partial oxidation (POX) reformer was demonstrated to supply a continuous power output of 50 W. This paper discusses some of the design paths chosen and challenges faced during the thermal integration of the stack and reformer in aiding the system startup and shutdown along with balance of plant and power management solutions. The package design, system capabilities, and test results of the prototype unit are presented.

  2. Effects of foundation conditions on plant design

    International Nuclear Information System (INIS)

    Ehasz, J.L.

    1975-01-01

    Design considerations for nuclear plant foundations are examined including site stability, bearing capacity and settlement, dynamic response, and structural modeling techniques for dynamic analysis. (U.S.)

  3. Conceptual capital-cost estimate and facility design of the Mirror-Fusion Technology Demonstration Facility

    International Nuclear Information System (INIS)

    1982-09-01

    This report contains contributions by Bechtel Group, Inc. to Lawrence Livermore National Laboratory (LLNL) for the final report on the conceptual design of the Mirror Fusion Technology Demonstration Facility (TDF). Included in this report are the following contributions: (1) conceptual capital cost estimate, (2) structural design, and (3) plot plan and plant arrangement drawings. The conceptual capital cost estimate is prepared in a format suitable for inclusion as a section in the TDF final report. The structural design and drawings are prepared as partial inputs to the TDF final report section on facilities design, which is being prepared by the FEDC

  4. Spherical tokamak power plant design issues

    International Nuclear Information System (INIS)

    Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

    2000-01-01

    The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

  5. Demonstration of direct internal reforming for MCFC power plants

    Energy Technology Data Exchange (ETDEWEB)

    Aasberg-Petersen, K.; Christensen, P.S.; Winther, S.K. [HALDOR TOPSOE A/S, Lynby (Denmark)] [and others

    1996-12-31

    The conversion of methane into hydrogen for an MCFC by steam reforming is accomplished either externally or internally in the stack. In the case of external reforming the plant electrical efficiency is 5% abs. lower mainly because more parasitic power is required for air compression for stack cooling. Furthermore, heat produced in the stack must be transferred to the external reformer to drive the endothermic steam reforming reaction giving a more complex plant lay-out. A more suitable and cost effective approach is to use internal steam reforming of methane. Internal reforming may be accomplished either by Indirect Internal Reforming (DIR) and Direct Internal Reforming (DIR) in series or by DIR-only as illustrated. To avoid carbon formation in the anode compartment higher hydrocarbons in the feedstock are converted into hydrogen, methane and carbon oxides by reaction with steam in ail adiabatic prereformer upstream the fuel cell stack. This paper discusses key elements of the desire of both types of internal reforming and presents data from pilot plants with a combined total of more than 10,000 operating hours. The project is being carried out as part of the activities of the European MCFC Consortium ARGE.

  6. ESBWR-An economical passive plant design

    International Nuclear Information System (INIS)

    Gonzalez Lopez, A.; Rao, A.

    1996-01-01

    This paper provides an overview of the design features of the European Simplified Boiling Water Reactor (ESBWR) design. The ESBWR is a plant design that builds on the Simplified Boiling Water Reactor (SBWR) design described in Reference 1 and 2. The major objective of the ESBWR programme is to develop a plant design that utilizes the basic simplicity of the SBWR design features to improve overall economics as discussed in Reference 3. The design is being developed by an international team of utilities, designers and researchers, with the objective of applying it to the utility and regulatory requirements of Europe. (Author)

  7. The Continuous Wave Deuterium Demonstrator (CWDD) design and status

    Energy Technology Data Exchange (ETDEWEB)

    Todd, A.M.M. [Grumman Space and Electronics Corp., Princeton, NJ (United States); Nightingale, M.P.S. [AEA Industrial Technology, Culham (United Kingdom); Yule, T.J. [Argonne National Lab., IL (United States)

    1992-12-31

    The design of the Continuous Wave Deuterium Demonstrator (CWDD) and the status of the fabricated hardware is presented. The CWDD is a high brightness, 352 MHz, CW linear accelerator designed to deliver a 7.54 MeV, 80 mA D{sup {minus}} beam at a transverse normalized rms emittance of 0.11 {pi} mm-mrad and a longitudinal rms emittance of 0.20 {pi} mm-mrad. End-to-end beam dynamics analysis for nominal and off-design conditions is described. The tuning and predicted operational performance os the as-built device are also discussed. These results all indicate that the present design can meet the output performance specifications in the presence of combined errors at the limits of the specified engineering tolerances. Preliminary injector operations have been conducted at AEA Technologies, Culham Laboratory and at Argonne National Laboratory, where the CWDD is sited. Initial RGQ beam experiments at Argonne are projected for October 1993. DTL installation and commissioning will be completed in 1994.

  8. 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

  9. Design quality assurance for nuclear power plants

    International Nuclear Information System (INIS)

    1986-07-01

    This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig

  10. Design quality assurance for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-07-01

    This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig.

  11. Design of a uranium recovery pilot plant

    International Nuclear Information System (INIS)

    1984-01-01

    The engineering design of a pilot plant of uranium recover, is presented. The diagrams and specifications of the equipments such as pipelines, pumps, values tanks, filters, engines, etc... as well as metallic structure and architetonic design is also presented. (author)

  12. Design of plant safety model in plant enterprise engineering environment

    International Nuclear Information System (INIS)

    Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-01-01

    Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured

  13. Light and Plants. A Series of Experiments Demonstrating Light Effects on Seed Germination, Plant Growth, and Plant Development.

    Science.gov (United States)

    Downs, R. J.; And Others

    A brief summary of the effects of light on plant germination, growth and development, including photoperiodism and pigment formation, introduces 18 experiments and demonstrations which illustrate aspects of these effects. Detailed procedures for each exercise are given, the expected results outlined, and possible sources of difficulty discussed.…

  14. Study on design method for seismically isolated FBR plants

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Yabana, Shuichi; Ohtori, Yasuki; Ishida, Katsuhiko; Sawada, Yoshihiro; Shiojiri; Hiroo; Mazda, Taiji

    1998-01-01

    CRIEPI conducted 'Demonstration test on FBR seismic isolation system' from 1987 to 1996 under contract with Ministry of International Trade and Industry, Japan. In the demonstration test, base isolation technologies are prepared and demonstrated to apply to FBR and the design guidelines are proposed. In this report overall contents of the design guidelines entitled Design guidelines for seismically base isolated FBR plants' are included. The design guidelines, as a rule, are limited to apply to FBR plants where entire reactor building is isolated in the horizontal direction using laminated rubber bearings as isolators. The design guidelines and its concepts, however, will be useful for the development of similar guidelines for other isolation systems using different type of isolation methods and other nuclear facilities. The design guidelines consist of three parts and appendices. The first part is 'Policy for Safety Design of Base Isolated FBR Plants' specifying the principles and the requirements in the planning and the design for the safety of base isolated FBR plants. The second part is Policy for Seismic Design of Base Isolated FBR' describing the principles and the requirements in the seismic design and the evaluation of safety for base isolated FBR plants. The third part is 'Design Methods for Seismic Isolated FBR Plants' detailing the methods, procedures and parameters to be used in the design and the evaluation of safety fro base isolated FBR plants. In appendices examples of design procedures for base isolated reactor building and laminated rubber bearings as well as various test data on laminated rubber bearings, etc. are shown. (author)

  15. Thermal power plant design and operation

    CERN Document Server

    Sarkar, Dipak

    2015-01-01

    Thermal Power Plant: Design and Operation deals with various aspects of a thermal power plant, providing a new dimension to the subject, with focus on operating practices and troubleshooting, as well as technology and design. Its author has a 40-long association with thermal power plants in design as well as field engineering, sharing his experience with professional engineers under various training capacities, such as training programs for graduate engineers and operating personnel. Thermal Power Plant presents practical content on coal-, gas-, oil-, peat- and biomass-fueled thermal power

  16. Design Of Feedforward Controllers For Multivariable Plants

    Science.gov (United States)

    Seraji, Homayoun

    1989-01-01

    Controllers based on simple low-order transfer functions. Mathematical criteria derived for design of feedforward controllers for class of multiple-input/multiple-output linear plants. Represented by simple low-order transfer functions, obtained without reconstruction of states of commands and disturbances. Enables plant to track command while remaining unresponsive to disturbance in steady state. Feedback controller added independently to stabilize plant or to make control system less susceptible to variations in parameters of plant.

  17. ESBWR-an economical passive plant design

    International Nuclear Information System (INIS)

    Arnold, H.; Stoop, P.M.; Gonzales, A.; Rao, A.

    1996-01-01

    The ESBWR is a plant design that builds on the GKN Dodewaard natural-circulation reactor and the simplified boiling water reactor (SBWR) design. The major objective of the ESBWR program, which has been in place for the past 3 yr, is to develop a plant design with proven technology that improves the overall plant economics. It utilizes the experience and basic simplicity of the Dodewaard plant and 670-MW(electric) SBWR design features. The design is being developed by an international team of utilities, designers, and researchers. It is being designed to meet the utility and regulatory requirements of Europe. It also addresses the key economic challenges for future nuclear power stations

  18. Controller design for interval plants

    International Nuclear Information System (INIS)

    Al-Sunni, F.M.

    2003-01-01

    We make use of celebrated Kharitoniv theorem to come up with a design procedure for the stabilization of uncertain systems in the parameters using low order controllers. The proposed design is based on classical design methods. A Non-linear programming (NLP) approach for the design of higher order controllers is also presented. We present our results and give illustrating examples. (author)

  19. Designing the Cash and Counseling Demonstration and Evaluation.

    Science.gov (United States)

    Doty, Pamela; Mahoney, Kevin J; Simon-Rusinowitz, Lori

    2007-02-01

    The Cash and Counseling Demonstration and Evaluation (CCDE) was designed as an experiment in shifting the paradigm in home and community-based long-term care from a professional/bureaucratic model of service delivery to one emphasizing consumer choice and control. The experimental intervention was an individualized budget offered in lieu of traditional Medicaid-covered services, such as agency-delivered aide services or a plan of care developed and coordinated by a professional case-manager, which typically involves authorization for several different providers to deliver a range of services. Within the spending limits established by their budgets, program participants were largely free to choose the types and amounts of paid services and supports they judged best able to meet their disability-related personal assistance needs. Medicaid beneficiaries in selected states who volunteered to participate. In all of the participating state Medicaid programs, beneficiaries eligible to participate included elders and younger adults with chronic disabilities and, in one state, adults and children with mental retardation/developmental disabilities could also participate. Minor children and adults with cognitive impairment could participate via representatives (family or friends who agreed to assist them in managing their services or to act as their surrogate decision-makers). Members of the CCDE management team describe the rationale for and implications of key design decisions. Key design decisions included the choice of research methodology (random assignment of CCDE participants in each state to treatment and control groups), selection of the state sites (AR, FL, NJ, NY), and the need for the CCDE to comply with federal waiver requirements for Medicaid research and demonstration projects. Principle Findings. The CCDE design was successfully implemented in three of the four state Medicaid programs selected for participation. The successful implementation of the CCDE

  20. Safety design of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

    1984-01-01

    Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

  1. Concept designs for NASA's Solar Electric Propulsion Technology Demonstration Mission

    Science.gov (United States)

    Mcguire, Melissa L.; Hack, Kurt J.; Manzella, David H.; Herman, Daniel A.

    2014-01-01

    Multiple Solar Electric Propulsion Technology Demonstration Mission were developed to assess vehicle performance and estimated mission cost. Concepts ranged from a 10,000 kilogram spacecraft capable of delivering 4000 kilogram of payload to one of the Earth Moon Lagrange points in support of future human-crewed outposts to a 180 kilogram spacecraft capable of performing an asteroid rendezvous mission after launched to a geostationary transfer orbit as a secondary payload. Low-cost and maximum Delta-V capability variants of a spacecraft concept based on utilizing a secondary payload adapter as the primary bus structure were developed as were concepts designed to be co-manifested with another spacecraft on a single launch vehicle. Each of the Solar Electric Propulsion Technology Demonstration Mission concepts developed included an estimated spacecraft cost. These data suggest estimated spacecraft costs of $200 million - $300 million if 30 kilowatt-class solar arrays and the corresponding electric propulsion system currently under development are used as the basis for sizing the mission concept regardless of launch vehicle costs. The most affordable mission concept developed based on subscale variants of the advanced solar arrays and electric propulsion technology currently under development by the NASA Space Technology Mission Directorate has an estimated cost of $50M and could provide a Delta-V capability comparable to much larger spacecraft concepts.

  2. Development of a Field Demonstration for Cost-Effective Low-Grade Heat Recovery and Use Technology Designed to Improve Efficiency and Reduce Water Usage Rates for a Coal-Fired Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Noble, Russell [Southern Company Services, Incorporated, Birmingham, AL (United States); Dombrowski, K. [AECOM Technical Services, Austin, TX (United States); Bernau, M. [AECOM Technical Services, Austin, TX (United States); Morett, D. [AECOM Technical Services, Austin, TX (United States); Maxson, A. [EPRI, Palo Alto, CA (United States); Hume, S. [EPRI, Palo Alto, CA (United States)

    2016-06-30

    Coal-based power generation systems provide reliable, low-cost power to the domestic energy sector. These systems consume large amounts of fuel and water to produce electricity and are the target of pending regulations that may require reductions in water use and improvements in thermal efficiency. While efficiency of coal-based generation has improved over time, coal power plants often do not utilize the low-grade heat contained in the flue gas and require large volumes of water for the steam cycle make-up, environmental controls, and for process cooling and heating. Low-grade heat recovery is particularly challenging for coal-fired applications, due in large part to the condensation of acid as the flue gas cools and the resulting potential corrosion of the heat recovery materials. Such systems have also not been of significant interest as recent investments on coal power plants have primarily been for environmental controls due to more stringent regulations. Also, in many regions, fuel cost is still a pass-through to the consumer, reducing the motivation for efficiency improvements. Therefore, a commercial system combining low-grade heat-recovery technologies and associated end uses to cost effectively improve efficiency and/or reduce water consumption has not yet been widely applied. However, pressures from potential new regulations and from water shortages may drive new interest, particularly in the U.S. In an effort to address this issue, the U.S. Department of Energy (DOE) has sought to identify and promote technologies to achieve this goal.

  3. Shielding design study of the demonstration fast breeder reactor. 2. Shielding design on the basis of the JASPER analysis

    International Nuclear Information System (INIS)

    Suzuoki, Zenro; Tabayashi, Masao; Handa, Hiroyuki; Iida, Masaaki; Takemura, Morio

    2000-01-01

    Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR. (author)

  4. BACA Project: geothermal demonstration power plant. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    The various activities that have been conducted by Union in the Redondo Creek area while attempting to develop the resource for a 50 MW power plant are described. The results of the geologic work, drilling activities and reservoir studies are summarized. In addition, sections discussing the historical costs for Union's involvement with the project, production engineering (for anticipated surface equipment), and environmental work are included. Nineteen geothermal wells have been drilled in the Redondo Creek area of the Valles Caldera: a prominent geologic feature of the Jemez mountains consisting of Pliocene and Pleistocene age volcanics. The Redondo Creek area is within a complex longitudinal graben on the northwest flank of the resurgent structural dome of Redondo Peak and Redondo Border. The major graben faults, with associated fracturing, are geologically plausible candidates for permeable and productive zones in the reservoir. The distribution of such permeable zones is too erratic and the locations too imprecisely known to offer an attractive drilling target. Log analysis indicates there is a preferred mean fracture strike of N31W in the upper portion of Redondo Creek wells. This is approximately perpendicular to the major structure in the area, the northeast-striking Redondo Creek graben. The geothermal fluid found in the Redondo Creek reservoir is relatively benign with low brine concentrations and moderate H/sub 2/S concentrations. Geothermometer calculations indicate that the reservoir temperature generally lies between 500/sup 0/F and 600/sup 0/F, with near wellbore flashing occurring during the majority of the wells' production.

  5. Hanford Waste Vitrification Plant technical background document for best available radionuclide control technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, A.B.; Skone, S.S.; Rodenhizer, D.G.; Marusich, M.V. (Ebasco Services, Inc., Bellevue, WA (USA))

    1990-10-01

    This report provides the background documentation to support applications for approval to construct and operate new radionuclide emission sources at the Hanford Waste Vitrification Plant (HWVP) near Richland, Washington. The HWVP is required to obtain permits under federal and state statutes for atmospheric discharges of radionuclides. Since these permits must be issued prior to construction of the facility, draft permit applications are being prepared, as well as documentation to support these permits. This report addresses the applicable requirements and demonstrates that the preferred design meets energy, environmental, and economic criteria for Best Available Radionuclide Control Technology (BARCT) at HWVP. 22 refs., 11 figs., 25 tabs.

  6. Hanford Waste Vitrification Plant technical background document for best available radionuclide control technology demonstration

    International Nuclear Information System (INIS)

    Carpenter, A.B.; Skone, S.S.; Rodenhizer, D.G.; Marusich, M.V.

    1990-10-01

    This report provides the background documentation to support applications for approval to construct and operate new radionuclide emission sources at the Hanford Waste Vitrification Plant (HWVP) near Richland, Washington. The HWVP is required to obtain permits under federal and state statutes for atmospheric discharges of radionuclides. Since these permits must be issued prior to construction of the facility, draft permit applications are being prepared, as well as documentation to support these permits. This report addresses the applicable requirements and demonstrates that the preferred design meets energy, environmental, and economic criteria for Best Available Radionuclide Control Technology (BARCT) at HWVP. 22 refs., 11 figs., 25 tabs

  7. Waste water pilot plant research, development, and demonstration permit application

    International Nuclear Information System (INIS)

    1991-10-01

    Waste waters have been generated as result of operations conducted at the Hanford Facility for over 40 years. These waste waters were previously discharged to cribs, ponds, or ditches. Examples of such waste waters include steam condensates and cooling waters that have not been in contact with dangerous or mixed waste and process condensates that may have been in contact with dangerous or mixed waste. Many measures have been taken to reduce the amount of contamination being discharged in these effluents. However, some of these waste waters still require additional treatment before release to the environment. Systems are being designed and built to treat these waste waters along with any future waste waters resulting from remediation activities on the Hanford Facility

  8. Information management systems improve advanced plant design

    International Nuclear Information System (INIS)

    Turk, R.S.; Serafin, S.A.; Leckley, J.B.

    1994-01-01

    Computer-aided engineering tools are proving invaluable in both the design and operation of nuclear power plants. ABB Combustion Engineering's Advanced Light Water Reactor (ALWR) features a computerized Information Management System (IMS) as an integral part of the design. The System 80+IMS represents the most powerful information management tool for Nuclear Power Plants commercially available today. Developed by Duke Power Company specifically for use by nuclear power plant owner operators, the IMS consists of appropriate hardware and software to manage and control information flow for all plant related work or tasks in a systematic, consistent, coordinated and informative manner. A significant feature of this IMS is that it is primarily based on plant data. The principal design tool, PASCE (Plant Application and Systems from Combustion Engineering), is comprised of intelligent databases that describe the design and from which accurate plant drawings are created. Additionally the IMS includes, at its hub, a relational database management system and an associated document management system. The data-based approach and applications associated with the IMS were developed, and have proven highly effective, for plant modifications, configuration management, and operations and maintenance applications at Duke Power Company's operating nuclear plants. This paper presents its major features and benefits. 4 refs

  9. MHI - Westinghouse joint FBR tank plant design

    International Nuclear Information System (INIS)

    Arnold, W.H.; Vijuk, R.M.; Aoki, I.; Messhil, T.

    1988-01-01

    Mitsubishi Heavy Industries and Westinghouse Advanced Energy Systems Division have combined their experience and capabilities to design a tank type fast breeder reactor plant. This tank type reactor has been refined and improved during the last three years to better compete in cost, safety, and operation with alternative power plants. This Mitsubishi/Westinghouse joint design offers economic advantages due to the use of steel structures, modular construction, nitrogen cells for the intermediate loops, reactor cavity air cooling and the use of the guard vessel as the containment vessel. Inherent characteristics in the reactor design provide protection to the public and the plant investment

  10. Virtual environments for nuclear power plant design

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W.

    1996-01-01

    In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP)

  11. Tritium supply assessment for ITER and DEMOnstration power plant

    International Nuclear Information System (INIS)

    Ni, Muyi; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-01-01

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource

  12. Tritium supply assessment for ITER and DEMOnstration power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Muyi, E-mail: muyi.ni@fds.org.cn; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-10-15

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource.

  13. Preparation for full scale demonstration of an air staged gasifier plant. Technical project development; For combined heat and power production with wood chips; Forberedelse til fuldskala demonstration af trinopdelt forgasningsanlaeg. Teknisk projektudvikling. Delrapport

    Energy Technology Data Exchange (ETDEWEB)

    Houmann Jakobsen, H.

    2011-04-15

    The project has aimed to further develop the technology for staged biomass gasification and establish an organizational and financial model to ensure that the technology can be introduced on the market. This report describes the technique in an upcoming demonstration plant. A complete planning and design of a demonstration plant with a capacity of 300 kW electric power and 700 kW heat was prepared. That is four times more than the pilot plant at Graested District Heating (Castor plant) can produce. A full scale demonstration plant with bio-gasification technology for wood chips will be established and put into operation in 2012. (ln)

  14. The ESDRED project: Engineering studies and demonstration of repository designs

    International Nuclear Information System (INIS)

    Verstricht, J.

    2009-01-01

    The construction, operation and closure of a deep geological repository for spent fuel and long-lived radioactive waste in clay involves specific technologies. The demonstration of these techniques at an industrial scale is being carried out in the frame of a technological integrated project within the sixth Framework Programme of EURATOM. The Belgian design for high level waste disposal is based on the so-called Supercontainer concept. Within this concept, the waste is encased in a carbon steel overpack, which is consequently fitted into a 70 cm thick concrete shell, in its turn enveloped by a stainless steel liner. A Supercontainer measures about 2 m in diameter. In the design of the repository, the Supercontainers will be emplaced, one after the other, in disposal galleries. The space between the Supercontainers and the gallery lining needs to be filled up with a solid material. The most essential function of this component, referred to as backfill, is to prevent a collapse of the gallery. A secondary function is to limit the presence of free oxygen, to limit corrosion. In the ESDRED project EIG EURIDICE, together with SCK-CEN and ONDRAF/NIRAS, investigates technologies to apply the backfill. After testing two techniques to apply the backfill in 2007 at limited scale (unite with granular material and grouting with backfill mortar), grouting was selected as the preferred technique. This technique then should be tested at full-scale (30 m long mock-up). First, a full-scale structure needs to built, including an extensive instrumentation programme. In addition, the logistical needs to ensure a continuous backfill operation have to be worked out. The objective is to have the almost 100 m 3 backfilled in 4 hours

  15. Secure, Mobile, Wireless Network Technology Designed, Developed, and Demonstrated

    Science.gov (United States)

    Ivancic, William D.; Paulsen, Phillip E.

    2004-01-01

    The inability to seamlessly disseminate data securely over a high-integrity, wireless broadband network has been identified as a primary technical barrier to providing an order-of-magnitude increase in aviation capacity and safety. Secure, autonomous communications to and from aircraft will enable advanced, automated, data-intensive air traffic management concepts, increase National Air Space (NAS) capacity, and potentially reduce the overall cost of air travel operations. For the first time ever, secure, mobile, network technology was designed, developed, and demonstrated with state-ofthe- art protocols and applications by a diverse, cooperative Government-industry team led by the NASA Glenn Research Center. This revolutionary technology solution will make fundamentally new airplane system capabilities possible by enabling secure, seamless network connections from platforms in motion (e.g., cars, ships, aircraft, and satellites) to existing terrestrial systems without the need for manual reconfiguration. Called Mobile Router, the new technology autonomously connects and configures networks as they traverse from one operating theater to another. The Mobile Router demonstration aboard the Neah Bay, a U.S. Coast Guard vessel stationed in Cleveland, Ohio, accomplished secure, seamless interoperability of mobile network systems across multiple domains without manual system reconfiguration. The Neah Bay was chosen because of its low cost and communications mission similarity to low-Earth-orbiting satellite platforms. This technology was successfully advanced from technology readiness level (TRL) 2 (concept and/or application formation) to TRL 6 (system model or prototype demonstration in a relevant environment). The secure, seamless interoperability offered by the Mobile Router and encryption device will enable several new, vehicle-specific and systemwide technologies to perform such things as remote, autonomous aircraft performance monitoring and early detection and

  16. Design concepts of nuclear desalination plants

    International Nuclear Information System (INIS)

    2002-11-01

    Interest in using nuclear energy for producing potable water has been growing worldwide in the past decade. This has been motivated by a variety of factors, including economic competitiveness of nuclear energy, the growing need for worldwide energy supply diversification, the need to conserve limited supplies of fossil fuels, protecting the environment from greenhouse gas emissions, and potentially advantageous spin-off effects of nuclear technology for industrial development. Various studies, and at least one demonstration project, have been considered by Member States with the aim of assessing the feasibility of using nuclear energy for desalination applications under specific conditions. In order to facilitate information exchange on the subject area, the IAEA has been active for a number of years in compiling related technical publications. In 1999, an inter regional technical co-operation project on Integrated Nuclear Power and desalination System Design was launched to facilitate international collaboration for the joint development by technology holders and potential end users of an integrated nuclear desalination system. This publication presents material on the current status of nuclear desalination activities and preliminary design concepts of nuclear desalination plants, as made available to the IAEA by various Member States. It is aimed at planners, designers and potential end-users in those Member States interested in further assessment of nuclear desalination. Interested readers are also referred to two related and recent IAEA publications, which contain useful information in this area: Introduction of Nuclear Desalination: A Guidebook, Technical Report Series No. 400 (2000) and Safety Aspects of Nuclear Plants Coupled with Seawater Desalination Units, IAEA-TECDOC-1235 (2001)

  17. Current fusion power plant design concepts

    International Nuclear Information System (INIS)

    Gore, B.F.; Murphy, E.S.

    1976-09-01

    Nine current U.S. designs for fusion power plants are described in this document. Summary tabulations include a tenth concept, for which the design document was unavailable during preparation of the descriptions. The information contained in the descriptions was used to define an envelope of fusion power plant characteristics which formed the basis for definition of reference first commercial fusion power plant design. A brief prose summary of primary plant features introduces each of the descriptions contained in the body of this document. In addition, summary tables are presented. These tables summarize in side-by-side fashion, plant parameters, processes, combinations of materials used, requirements for construction materials, requirements for replacement materials during operation, and production of wastes

  18. The AX-PET demonstrator-Design, construction and characterization

    International Nuclear Information System (INIS)

    Beltrame, P.; Bolle, E.; Braem, A.; Casella, C.; Chesi, E.; Clinthorne, N.; De Leo, R.; Dissertori, G.; Djambazov, L.; Fanti, V.; Heller, M.; Joram, C.; Kagan, H.; Lustermann, W.; Meddi, F.; Nappi, E.; Nessi-Tedaldi, F.; Oliver, J.F.; Pauss, F.; Rafecas, M.

    2011-01-01

    Axial PET is a novel geometrical concept for Positron Emission Tomography (PET), based on layers of long scintillating crystals axially aligned with the bore axis. The axial coordinate is obtained from arrays of wavelength shifting (WLS) plastic strips placed orthogonally to the crystals. This article describes the design, construction and performance evaluation of a demonstrator set-up which consists of two identical detector modules, used in coincidence. Each module comprises 48 LYSO crystals of 100 mm length and 156 WLS strips. Crystals and strips are readout by Geiger-mode Avalanche Photo Diodes (G-APDs). The signals from the two modules are processed by fully analog front-end electronics and recorded in coincidence by a VME-based data acquisition system. Measurements with point-like 22 Na sources, with the modules used both individually and in coincidence mode, allowed for a complete performance evaluation up to the focal plane reconstruction of point sources. The results obtained are in good agreement with expectations and proved the set-up to be ready for the next evaluation phase with PET phantoms filled with radiotracers.

  19. Design demonstrations for Category B tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1992-07-01

    This document presents design demonstrations conducted of liquid low-level waste (LLLW) storage tank systems located at the Oak Ridge National Laboratory (ORNL). ORNL has conducted research in energy related fields since 1943. The facilities used to conduct the research include nuclear reactors, chemical pilot plants, research laboratories, radioisotope production laboratories, and support facilities. These facilities have produced a variety of radioactive and/or hazardous wastes. These wastes have been stored and transported through an extensive network of piping and tankage. Demonstration of the design of these tank systems has been stipulated by the Federal Facility Agreement (FFA) between the US Environmental Protection Agency (EPA) - Region IV; the Tennessee Department of Environment and Conservation (TDEC); and the DOE. The FFA establishes four categories of tanks. These are: Category A -- New or Replacement Tank Systems with Secondary Containment; Category B -- Existing Tank Systems with Secondary Containment; Category C -- Existing Tank Systems without Secondary Containment; and Category D -- Existing Tank Systems without Secondary Containment that are; Removed from Service. This document provides a design demonstration of the secondary containment and ancillary equipment of 11 tank systems listed in the FFA as Category ''B.'' The design demonstration for each tank is presented in Section 2. The design demonstrations were developed using information obtained from the design drawings (as-built when available), construction specifications, and interviews with facility operators. The assessments assume that each tank system was constructed in accordance with the design drawings and construction specifications for that system unless specified otherwise. Each design demonstration addresses system conformance to the requirements of the FFA (Appendix F, Subsection C)

  20. IRIS Nuclear Power Plant design

    International Nuclear Information System (INIS)

    Carelli, M. D.; Cobian, J.

    2002-01-01

    IRIS(International Reactor Innovative and Secure) is a novel light water reactor with a modular, integral primary system configuration. This concept, initially developed in response to the first NERI solicitation, is now being pursued by an international consortium of 20 participants from seven countries. IRIS is designed to satisfy the four key requirements for Generation IV systems: enhanced safety, improved economics, proliferation resistance and waste minimization. Its main features are: small-to-medium power (100-335 MWe/module); long life core 5 to 10 years) without shuffling or refueling; optimized maintenance with repair shutdown intervals of a least four years; simplified compact design with the primary vessel housing steam generators, pressurizer and pumps; safety by design where accidents are positively eliminated by design rather than engineering to cope with their consequences; loss of coolant accidents of any size and loss of low accidents are eliminated as major safety concerns; estimated power generation total cost is projected to be competitive with other power options. IRIS is one of four new reactor designs currently under NRC review. Projected schedule calls for design certification by 2008 and being ready for deployment by 2001 or later. This rather short schedule is made possible by the fact that IRIS is based on proven light water technology and new technology development is not required. (Author)

  1. Draft plan for the Waste Isolation Pilot Plant test phase: Performance assessment and operations demonstration

    International Nuclear Information System (INIS)

    1989-04-01

    The mission of the Waste Isolation Pilot Plant (WIPP) Project is to provide a research and development facility to demonstrate the safe disposal of transuranic (TRU) radioactive wastes resulting from United States defense programs. With the Construction Phase of the WIPP facility nearing completion, WIPP is ready to initiate the next phase in its development, the Test Phase. The purpose of the Test Phase is to collect the necessary scientific and operational data to support a determination whether to proceed to the Disposal Phase and thereby designate WIPP a demonstration facility for the disposal of TRU wastes. This decision to proceed to the Disposal Phase is scheduled for consideration by September 1994. Development of the WIPP facility is the responsibility of the United States Department of Energy (DOE), whose Albuquerque Operations Office has designated the WIPP Project Office as Project Manager. This document describes the two major programs to be conducted during the Test Phase of WIPP: (1) Performance Assessment for determination of compliance with the Environmental Protection Agency Standard and (2) Operations Demonstration for evaluation of the safety and effectiveness of the DOE TRU waste management system's ability to emplace design throughput quantities of TRU waste in the WIPP facility. 42 refs., 38 figs., 14 tabs

  2. Conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR), September 1978

    International Nuclear Information System (INIS)

    Kelley, J.L.

    1978-12-01

    The flexibility of the fusion hybrid reactor to function as a fuel production facility, power plant, waste disposal burner or combinations of all of these, as well as the reactor's ability to use proliferation resistant fuel cycles, has provided the incentive to assess the feasibility of a near-term demonstration plant. The goals for a Demonstration Tokamak Hybrid Reactor (DTHR) were established and an initial conceptual design was selected. Reactor performance and economics were evaluated and key developmental issues were assessed. The study has shown that a DTHR is feasible in the late 1980's, a significant quantity of fissile fuel could be produced from fertile thorium using present day fission reactor blanket technology, and a large number of commercially prototypical components and systems could be developed and operationally verified. The DTHR concept would not only serve as proof-of-principle for hybrid technology, but could be operated in the ignited mode and provide major advancements for pure fusion technology

  3. Pre-engineering assessment of Enersolve Demonstration Project. Dairy processing plant. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-15

    This study involved evaluation of a dairy processing plant to demonstrate energy conservation potential and developing a strategy to realize energy savings through retrofitting and revamping. The cheese and whey making plant of Agropur Cooperative Agro-Alimentaire located in Quebec was selected as a representative Canadian dairy processing plant. The pre-engineering assessment included a review of existing facility at the plant and plant operation, identification of plant equipment or process steps where substantial economic benefits would result from retrofitting and revamping, and estimation of budgetary cost for the subsequent engineering, procurement, construction management and monitoring of the retrofitted equipment. 3 figs., 2 tabs.

  4. LBB considerations for a new plant design

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Mandava, P.R.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1997-04-01

    The leak-before-break (LBB) methodology is accepted as a technically justifiable approach for eliminating postulation of Double-Ended Guillotine Breaks (DEGB) in high energy piping systems. This is the result of extensive research, development, and rigorous evaluations by the NRC and the commercial nuclear power industry since the early 1970s. The DEGB postulation is responsible for the many hundreds of pipe whip restraints and jet shields found in commercial nuclear plants. These restraints and jet shields not only cost many millions of dollars, but also cause plant congestion leading to reduced reliability in inservice inspection and increased man-rem exposure. While use of leak-before-break technology saved hundreds of millions of dollars in backfit costs to many operating Westinghouse plants, value-impacts resulting from the application of this technology for future plants are greater on a per plant basis. These benefits will be highlighted in this paper. The LBB technology has been applied extensively to high energy piping systems in operating plants. However, there are differences between the application of LBB technology to an operating plant and to a new plant design. In this paper an approach is proposed which is suitable for application of LBB to a new plant design such as the Westinghouse AP600. The approach is based on generating Bounding Analyses Curves (BAC) for the candidate piping systems. The general methodology and criteria used for developing the BACs are based on modified GDC-4 and Standard Review Plan (SRP) 3.6.3. The BAC allows advance evaluation of the piping system from the LBB standpoint thereby assuring LBB conformance for the piping system. The piping designer can use the results of the BACs to determine acceptability of design loads and make modifications (in terms of piping layout and support configurations) as necessary at the design stage to assure LBB for the, piping systems under consideration.

  5. Control of hydrogen sulfide emission from geothermal power plants. Volume III. Final report: demonstration plant equipment descriptions, test plan, and operating instructions

    Energy Technology Data Exchange (ETDEWEB)

    Brown, F.C.; Harvey, W.W.; Warren, R.B.

    1977-01-01

    The elements of the final, detailed design of the demonstration plant for the copper sulfate process for the removal of hydrogen sulfide from geothermal steam are summarized. Descriptions are given of all items of equipment in sufficient detail that they can serve as purchase specifications. The process and mechanical design criteria which were used to develop the specifications, and the process descriptions and material and energy balance bases to which the design criteria were applied are included. (MHR)

  6. A study of hazardous air pollutants at the Tidd PFBC Demonstration Plant

    International Nuclear Information System (INIS)

    1994-10-01

    The US Department of Energy (DOE) Clean Coal Technology (CCD Program is a joint effort between government and industry to develop a new generation of coal utilization processes. In 1986, the Ohio Power Company, a subsidiary of American Electric Power (AEP), was awarded cofunding through the CCT program for the Tidd Pressure Fluidized Bed Combustor (PFBC) Demonstration Plant located in Brilliant, Ohio. The Tidd PFBC unit began operation in 1990 and was later selected as a test site for an advanced particle filtration (APF) system designed for hot gas particulate removal. The APF system was sponsored by the DOE Morgantown Energy Technology Center (METC) through their Hot Gas Cleanup Research and Development Program. A complementary goal of the DOE CCT and METC R ampersand D programs has always been to demonstrate the environmental acceptability of these emerging technologies. The Clean Air Act Amendments of 1990 (CAAA) have focused that commitment toward evaluating the fate of hazardous air pollutants (HAPs) associated with advanced coal-based and hot gas cleanup technologies. Radian Corporation was contacted by AEP to perform this assessment of HAPs at the Tidd PFBC demonstration plant. The objective of this study is to assess the major input, process, and emission streams at Plant Tidd for the HAPs identified in Title III of the CAAA. Four flue gas stream locations were tested: ESP inlet, ESP outlet, APF inlet, and APF outlet. Other process streams sampled were raw coal, coal paste, sorbent, bed ash, cyclone ash, individual ESP hopper ash, APF ash, and service water. Samples were analyzed for trace elements, minor and major elements, anions, volatile organic compounds, dioxin/furan compounds, ammonia, cyanide, formaldehyde, and semivolatile organic compounds. The particle size distribution in the ESP inlet and outlet gas streams and collected ash from individual ESP hoppers was also determined

  7. New design system for nuclear power plant

    International Nuclear Information System (INIS)

    Kakuta, Masataka; Yoshinaga, Toshiaki; Yoshida, Ikuzo; Tokumasu, Shinji.

    1980-01-01

    As for the machine and equipment layout and the piping design for nuclear power plants, the multilateral coordination and study on such factors as functions, installation, radiation exposure and maintenance are required, and the high reliability is demanded. On the other hand, the quantity of things handled is enormous, therefore it is difficult to satisfy completely the above described requirements and to make plant planning which is completely free from the mutual interference of machines, equipments and pipings by the ordinary design with drawings only. Thereupon, the following new device was adopted to the design method for the purposes of improving the quality and shortening the construction period. Namely at the time of designing new plants, the rationalization of plant planning method was attempted by introducing color composite drawings and the technique of model engineering, at the same time, the newly developed design system for pipings was applied with a computer, thus the large accomplishment was able to be obtained regarding the improvement of reliability and others by making the check-up of the propriety. The design procedures of layout and piping, the layout design and general coordination in nuclear power stations with models and color composite drawings and the design system are explained. (Kako, I.)

  8. SEPI an expert system for plant design

    International Nuclear Information System (INIS)

    Carotenuto, M.; Corleto, P.; Landeyro, P.

    1988-01-01

    The availability and suitability of technological information is of great importance in every kind of design task, especially when safety and reliability considerations are involved. In this paper an ''expert system for plant design'' (SEPI), is presented, together with its first application to nuclear back-end plants. This system is available on ENEA computer network. It is thought to be used both to collect know-how developed in the field and to assist unskilled designers during selection, evaluation and dimensioning tasks. It attemps to reproduce the normal way of ''reasoning'' and acting, and provides some graphic facilities

  9. TVA coal-gasification commercial demonstration plant project. Volume 5. Plant based on Koppers-Totzek gasifier. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-11-01

    This volume presents a technical description of a coal gasification plant, based on Koppers-Totzek gasifiers, producing a medium Btu fuel gas product. Foster Wheeler carried out a conceptual design and cost estimate of a nominal 20,000 TPSD plant based on TVA design criteria and information supplied by Krupp-Koppers concerning the Koppers-Totzek coal gasification process. Technical description of the design is given in this volume.

  10. Ames expedited site characterization demonstration at the former manufactured gas plant site, Marshalltown, Iowa

    International Nuclear Information System (INIS)

    Bevolo, A.J.; Kjartanson, B.H.; Wonder, J.D.

    1996-03-01

    The goal of the Ames Expedited Site Characterization (ESC) project is to evaluate and promote both innovative technologies (IT) and state-of-the-practice technologies (SOPT) for site characterization and monitoring. In April and May 1994, the ESC project conducted site characterization, technology comparison, and stakeholder demonstration activities at a former manufactured gas plant (FMGP) owned by Iowa Electric Services (IES) Utilities, Inc., in Marshalltown, Iowa. Three areas of technology were fielded at the Marshalltown FMGP site: geophysical, analytical and data integration. The geophysical technologies are designed to assess the subsurface geological conditions so that the location, fate and transport of the target contaminants may be assessed and forecasted. The analytical technologies/methods are designed to detect and quantify the target contaminants. The data integration technology area consists of hardware and software systems designed to integrate all the site information compiled and collected into a conceptual site model on a daily basis at the site; this conceptual model then becomes the decision-support tool. Simultaneous fielding of different methods within each of the three areas of technology provided data for direct comparison of the technologies fielded, both SOPT and IT. This document reports the results of the site characterization, technology comparison, and ESC demonstration activities associated with the Marshalltown FMGP site. 124 figs., 27 tabs

  11. Demonstration of a 1 MWe biomass power plant at USMC Base Camp Lejeune

    International Nuclear Information System (INIS)

    Cleland, J.; Purvis, C.R.

    1997-01-01

    A biomass energy conversion project is being sponsored by the U.S. Environmental Protection Agency (EPA) to demonstrate an environmentally and economically sound electrical power option for government installations, industrial sites, rural cooperatives, small municipalities, and developing countries. Under a cooperative agreement with EPA, Research Triangle Institute is initiating operation of the Camp Lejeune Energy from Wood (CLEW) biomass plant. Wood gasification combined with internal combustion engines was chosen because of (1) recent improvements in gas cleaning, (2) simple, economical operation for units less than 10 MW, and (3) the option of a clean, cheap fuel for the many existing facilities generating expensive electricity from petroleum fuels with reciprocating engines. The plant incorporates a downdraft, moving bed gasifier utilizing hogged waste wood from the Marine Corps Base at Camp Lejeune, NC. A moving bed bulk wood dryer and both spark ignition and diesel engines are included. Unique process design features are briefly described relative to the gasifier, wood drying, tar separation, and process control. A test plan for process optimization and demonstration of reliability, economics, and environmental impact is outlined. (author)

  12. Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

    International Nuclear Information System (INIS)

    1978-01-01

    This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified

  13. It is now time to proceed with a gas-cooled breeder reactor (GBR) demonstration plant

    International Nuclear Information System (INIS)

    Chermanne, J.

    1976-01-01

    Since 1969, the GBRA has been making studies to provide evidence on questions which were not clear regarding the Gas-cooled Breeder Reactor: design feasibility and performance, safety, strategy and economics, and R and D necessary for a demonstration plant. Studies were carried out on a 1200-MW(e) commercial reference design with pin fuel, which was also used as a basis for a definition of the GBR demonstration plant. During the six years, a great deal of information has been generated at GBRA and it confirms the forecasts of the promoters of the Gas-cooled Breeder Reactor that the GBR is an excellent reactor from all points of view: design - the reactor can be engineered without major difficulty, using present techniques. As far as fuel is concerned, LMFBR fuel technology is applicable plus limited specific development effort. Performance - the GBR is the best breeder with oxide fuel and using conventional techniques. The strategy studies carried out at GBRA have clearly shown that a high performance breeder such as the GBR is absolutely required in large quantities by the turn of the century in order to avoid dependence on natural uranium resources. Regarding safety, a major step forward has been made when an ad hoc group on GBR safety, sponsored by the EEC, concluded that no major difficulties were anticipated which would prevent the GBR reaching adequate safety standards. Detailed economic assessments performed on an item-to-item basis have shown that the cost of a GBR with its high safety standard is about the same as that of an HTR. One can therefore conclude that, with the present cost of natural uranium, the GBR is competitive with the LWRs. Owing to the very limited R and D effort necessary and the obvious safety, economic and strategic advantages of the concept, it is concluded that the development and construction of a GBR demonstration plant must be started now if one wants to secure an adequate energy supply past the turn of the century. (author)

  14. Considering plant life management influences on new plant design

    International Nuclear Information System (INIS)

    Dam, R.F.; Choy, E.; Soulard, M.; Nickerson, J.H.; Hopwood, J.

    2003-01-01

    After operating successfully for more than half their design life, owners of CANDU reactors are now engaging in Plant Life Management (PLiM) activities to ensure not only life attainment, but also life extension. For several years, Atomic Energy of Canada Ltd. (AECL) has been working with domestic and offshore CANDU utilities on a comprehensive and integrated CANDU PLiM program that will see existing CANDU plants successfully and reliably operate through their design life and beyond. To support the PLiM program development, a significant level of infrastructure has been, and continues to be, developed at AECL. This includes the development of databases that document relevant knowledge and background to allow for a more accessible and complete understanding of degradation issues and the strategies needed to deal with these issues. As the level of integration with various project, services and R and D activities in AECL increases, this infrastructure is growing to encompass a wider range of design, operations and maintenance details to support comprehensive and quantitative assessment of CANDU stations. With the maturation of the PLiM program, these processes were adapted for application to newer plants. In particular, a fully integrated program was developed that interrelates the design basis, operations, safety, and reliability and maintenance strategies, as applied to meet plant design goals. This has led to the development of the maintenance-based design concept. The various PLiM technologies, developed and applied in the above programs with operating stations, are being modified and tailored to assist with the new plant design processes to assure that ACR- Advanced CANDU Reactor meets its targets for operation, maintenance, and lifetime performance. Currently, the ACR, developed by Atomic Energy of Canada Ltd. (AECL), is being designed with features to increase capacity factors, to reduce the risk of major equipment failures, to improve access to key components

  15. Design in action: From prototyping by demonstration to cooperative prototyping

    DEFF Research Database (Denmark)

    Bødker, Susanne; Grønbæk, Kaj

    1991-01-01

    ... the development of any computer-based system will have to proceed in a cycle from design to experience and back again. It is impossible to anticipate all of the relevant breakdown and their domains. They emerge gradually in practice. Winograd and Flores, 1986. p.171 Some time ago we worked wi...... with a group of dental assistants, designing a prototype case record system to explore the possibility of using computer support in public dental clinics. ...

  16. Advanced power plant materials, design and technology

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, D. (ed.) [Newcastle University (United Kingdom). Sir Joseph Swan Institute

    2010-07-01

    The book is a comprehensive reference on the state of the art of gas-fired and coal-fired power plants, their major components and performance improvement options. Selected chapters are: Integrated gasification combined cycle (IGCC) power plant design and technology by Y. Zhu, and H. C. Frey; Improving thermal cycle efficiency in advanced power plants: water and steam chemistry and materials performance by B. Dooley; Advanced carbon dioxide (CO{sub 2}) gas separation membrane development for power plants by A. Basile, F. Gallucci, and P. Morrone; Advanced flue gas cleaning systems for sulphur oxides (SOx), nitrogen oxides (NOx) and mercury emissions control in power plants by S. Miller and B.G. Miller; Advanced flue gas dedusting systems and filters for ash and particulate emissions control in power plants by B.G. Miller; Advanced sensors for combustion monitoring in power plants: towards smart high-density sensor networks by M. Yu and A.K. Gupta; Advanced monitoring and process control technology for coal-fired power plants by Y. Yan; Low-rank coal properties, upgrading and utilisation for improving the fuel flexibility of advanced power plants by T. Dlouhy; Development and integration of underground coal gasification (UCG) for improving the environmental impact of advanced power plants by M. Green; Development and application of carbon dioxide (CO{sub 2}) storage for improving the environmental impact of advanced power plants by B. McPherson; and Advanced technologies for syngas and hydrogen (H{sub 2}) production from fossil-fuel feedstocks in power plants by P. Chiesa.

  17. Design of a nuclear steam reforming plant

    International Nuclear Information System (INIS)

    Malherbe, J.

    1980-01-01

    The design of a plant for the steam reforming of methane using a High Temperature Reactor has been studied by CEA in connection with the G.E.G.N. This group of companies (CEA, GAZ DE FRANCE, CHARBONNAGES DE FRANCE, CREUSOT-LOIRE, NOVATOME) is in charge of studying the feasibility of the coal gasification process by using a nuclear reactor. The process is based on the hydrogenation of the coal in liquid phase with hydrogen produced by a methane steam reformer. The reformer plant is fed by a pipe of natural gas or SNG. The produced hydrogen feeds the gasification plant which could not be located on the same site. An intermediate hydrogen storage between the two plants could make the coupling more flexible. The gasification plant does not need a great deal of heat and this heat can be satisfied mostly by internal heat exchanges

  18. Design and control of integrated styrene aniline production plant

    NARCIS (Netherlands)

    Partenie, O.; Van der Last, V.; Sorin Bildea, C.; Altimari, P.

    2009-01-01

    This paper illustrates the operational difficulties arising from simultaneously performing exothermic and endothermic reactions, and demonstrates that a plant can be built and safely operated by integrating the design and plantwide control issues. The behaviour of reactor – separation – recycle

  19. Demonstration testing and evaluation of in situ soil heating. Revision 1, Demonstration system design

    International Nuclear Information System (INIS)

    Dev, H.

    1994-01-01

    Over the last nine years IIT Research Institute (IITRI) has been developing and testing the in situ heating and soil decontamination process for the remediation of soils containing hazardous organic contaminants. In this process the soil is heated in situ using electrical energy. The contaminants are removed from the soil due to enhanced vaporization, steam distillation and stripping. The vaporized contaminants, water vapor and air are recovered from the heated zone by means of a vacuum manifold system which collects gases from below surface as well as from the soil surface. A vapor barrier is used to prevent fugitive emissions of the contaminants and to control air infiltration to minimize dilution of the contaminant gases and vapors. The recovered gases and vapors are conveyed to an on site vapor treatment system for the clean up of the vent gases. Electrical energy is applied to the soil by forming an array of electrodes in the soil which are electrically interconnected and supplied with power. The electrodes are placed in drilled bore holes which are made through the contaminated zone. There are two versions of the in situ heating and soil treatment process: the f irst version is called the In Situ Radio Frequency (RF) Soil Decontamination Process and the second version is called the In Situ Electromagnetic (EM) Soil Decontamination Process. The first version, the RF Process is capable of heating the soil in a temperature range of 100 degrees to 400 degrees C. The soil temperature in the second version, the EM Process, is limited to the boiling point of water under native conditions. Thus the soil will be heated to a temperature of about 85 degrees to 95 degrees C. In this project IITRI will demonstrate the EM Process for in situ soil decontamination at K-25 Site due to the fact that most of the contaminants of concern are volatile organics which can be removed by heating the soil to a temperature range of 85 degrees to 95 degrees C

  20. Demonstration testing and evaluation of in situ soil heating. Revision 1, Demonstration system design

    Energy Technology Data Exchange (ETDEWEB)

    Dev, H.

    1994-08-16

    Over the last nine years IIT Research Institute (IITRI) has been developing and testing the in situ heating and soil decontamination process for the remediation of soils containing hazardous organic contaminants. In this process the soil is heated in situ using electrical energy. The contaminants are removed from the soil due to enhanced vaporization, steam distillation and stripping. The vaporized contaminants, water vapor and air are recovered from the heated zone by means of a vacuum manifold system which collects gases from below surface as well as from the soil surface. A vapor barrier is used to prevent fugitive emissions of the contaminants and to control air infiltration to minimize dilution of the contaminant gases and vapors. The recovered gases and vapors are conveyed to an on site vapor treatment system for the clean up of the vent gases. Electrical energy is applied to the soil by forming an array of electrodes in the soil which are electrically interconnected and supplied with power. The electrodes are placed in drilled bore holes which are made through the contaminated zone. There are two versions of the in situ heating and soil treatment process: the f irst version is called the In Situ Radio Frequency (RF) Soil Decontamination Process and the second version is called the In Situ Electromagnetic (EM) Soil Decontamination Process. The first version, the RF Process is capable of heating the soil in a temperature range of 100{degrees} to 400{degrees}C. The soil temperature in the second version, the EM Process, is limited to the boiling point of water under native conditions. Thus the soil will be heated to a temperature of about 85{degrees} to 95{degrees}C. In this project IITRI will demonstrate the EM Process for in situ soil decontamination at K-25 Site due to the fact that most of the contaminants of concern are volatile organics which can be removed by heating the soil to a temperature range of 85{degrees} to 95{degrees}C.

  1. Demonstration of an automated on-line surveillance system at a commercial nuclear power plant

    International Nuclear Information System (INIS)

    Smith, C.M.; Sweeney, F.J.

    1983-01-01

    As a first step in demonstrating the practicality of performing continuous on-line surveillance of the performance of nuclear steam supply systems using noise related techniques, Oak Ridge National Laboratory is operating a computerized noise signal data acquisition and processing system at the Sequoyah Unit 1 Nuclear Plant, an 1148 MWe four-loop Westinghouse pressurized water reactor (PWR) located near Chattanooga, Tennessee. The principal objective is to establish, with a degree of continuity and completeness not previously achieved, the long-term characteristics of signals from neutron detectors and process sensors in order to evaluate the feasibility of detecting and diagnosing anomalous reactor conditions by means of changes in these signals. The system is designed to automatically screen the gathered data, using a number of descriptors derived from the power spectra of the monitored signals, and thereby select for the noise analyst's perusal only those data which differ statistically from norms which the system has previously established

  2. System 80+ Design and Licensing : Improving Plant Reliability

    International Nuclear Information System (INIS)

    Newman, Robert E.

    1989-01-01

    The U. S. nuclear industry is striving to improve plant reliability and availability through improved plant design, component designs and plant maintenance. In an effort to improve safety and to demonstrate that commercial nuclear power is economically competitive with other energy sources, the utilities, nuclear vendors, architect engineers and constructors, and component suppliers are all participating in an industry-wide effort to develop improved Light Water Reactor (LWR) designs that are based upon the many years of successful LWR operation. In an age when the world faces the environmental pressures of the greenhouse effect and acid rain, electricity generated from nuclear energy must play an increasing role in the energy picture of Korea, the United States and the rest of the world. This paper discusses the plant availability requirement that has been established by the industry-wide effort mentioned above. After briefly describing Combustion Engineering's program for development of the System 80 Plus standard design and the participation of the Korea Advanced Energy Research Institute (KAERI) in the program, the paper then describes the design features that are being incorporated into System 80+. The industry ALRR Program has established a very ambitious criterion of 87% for the plant availability of future nuclear units. To satisfy such a requirement, the next generation of nuclear plants will include a great many design improvements that reflect the hundreds of years of operating experience that we have accrued. C-ESA's System 80+ will include a number of design changes that improve operating margins and make the plant easier to operate and maintain. Not surprisingly, there is a great deal of overlap between improved safety and improved reliability. In the end, our design will satisfy the future needs of the utilities, the regulators, and the public. C-E is very pleased that KAERI is working with US to achieve these important goals

  3. Inferences from new plant design from fast flux test facility operation

    International Nuclear Information System (INIS)

    Peterson, R.E.; Peckinpaugh, C.L.; Simpson, D.E.

    1985-04-01

    Experience gained through operation of the Fast Flux Test Facility (FFTF) is now sufficiently extensive that this experience can be utilized in designing the next generation of liquid metal fast reactors. Experience with FFTF core and plant components is cited which can result in design improvements to achieve inherently safe, economic reactor plants. Of particular interest is the mixed oxide fuel system which has demonstrated large design margins. Other plant components have also demonstrated high reliability and offer capital cost reduction opportunities through design simplifications. The FFTF continues to be a valuable US resource which affords prototypic development and demonstration, contributing to public acceptability of future plants

  4. Development and demonstration of near-real-time accounting systems for reprocessing plants

    International Nuclear Information System (INIS)

    Cobb, D.D.; Hakkila, E.A.; Dayem, H.A.; Shipley, J.P.; Baker, A.L.

    1981-01-01

    A program to develop and demonstrate near-real-time accounting systems for reprocessing plants has been active at Los Alamos since 1976. The technology has been developed through modeling and simulation of process operation and measurement systems and evaluation of these data using decision analysis techniques. Aspects of near-real-time systems have been demonstrated successfully at the AGNS reprocessng plant as part of a joint study of near-real-time accounting

  5. Optimized solar heat production in a liberalised electricity market. Demonstration of full-scale plant in Braedstrup; Optimeret solvarmeproduktion i et liberaliseret elmarked. Demonstration af fuldskalaanlaeg i Braedstrup

    Energy Technology Data Exchange (ETDEWEB)

    Soerensen, P.A. (PlanEnergi, Skoerping (Denmark)); Kristensen, Per (Braedstrup Fjernvarme, Braedstrup (Denmark)); Furbo, S. (Danmarks Tekniske Univ. DTU BYG, Kgs. Lyngby (Denmark)); Ulbjerg, F. (Ramboell, Odense (Denmark)); Holm, L. (Marstal Fjernvarme, Marstal (Denmark)); Schmidt, T. (Steinbeis-Research Institute for Solar and Sustainable Thermal Systems, Stuttgart (Denmark))

    2009-03-15

    The project demonstrates for the first time a combination between CHP and solar power systems. 8,019 m2 solar collectors producing 8% of the annual consumption in Braedstrup, Denmark, and nearly the total consumption on a good summer day were combined with a natural gas-fired CHP plant. An optimised ARCON HT2006 collector was developed for this purpose, and the control system was designed to ensure that supply-pipe temperature from solar collectors is always as low as possible and that the temperature in the existing water storage tank does not drop below 90 deg. C. (ln)

  6. Remote maintenance demonstration tests at a pilot plant for high level waste vitrification

    International Nuclear Information System (INIS)

    Selig, M.

    1984-01-01

    The remote maintenance and replacement technique designed for a radioactive vitrification plant have been developed and tested in a full scale handling mockup and in an inactive pilot plants by the Central Engineering Department of the Karlsruhe Nuclear Research Center. As a result of the development work and the tests it has been proved that the remote maintenance technique and remote handling equipment can be used without any technical problems and are suited for application in a radioactive waste vitrification plant

  7. Phylogenomic analysis demonstrates a pattern of rare and ancient horizontal gene transfer between plants and fungi.

    Science.gov (United States)

    Richards, Thomas A; Soanes, Darren M; Foster, Peter G; Leonard, Guy; Thornton, Christopher R; Talbot, Nicholas J

    2009-07-01

    Horizontal gene transfer (HGT) describes the transmission of genetic material across species boundaries and is an important evolutionary phenomenon in the ancestry of many microbes. The role of HGT in plant evolutionary history is, however, largely unexplored. Here, we compare the genomes of six plant species with those of 159 prokaryotic and eukaryotic species and identify 1689 genes that show the highest similarity to corresponding genes from fungi. We constructed a phylogeny for all 1689 genes identified and all homolog groups available from the rice (Oryza sativa) genome (3177 gene families) and used these to define 14 candidate plant-fungi HGT events. Comprehensive phylogenetic analyses of these 14 data sets, using methods that account for site rate heterogeneity, demonstrated support for nine HGT events, demonstrating an infrequent pattern of HGT between plants and fungi. Five HGTs were fungi-to-plant transfers and four were plant-to-fungi HGTs. None of the fungal-to-plant HGTs involved angiosperm recipients. These results alter the current view of organismal barriers to HGT, suggesting that phagotrophy, the consumption of a whole cell by another, is not necessarily a prerequisite for HGT between eukaryotes. Putative functional annotation of the HGT candidate genes suggests that two fungi-to-plant transfers have added phenotypes important for life in a soil environment. Our study suggests that genetic exchange between plants and fungi is exceedingly rare, particularly among the angiosperms, but has occurred during their evolutionary history and added important metabolic traits to plant lineages.

  8. Conceptual Design of an Electric Sail Technology Demonstration Mission Spacecraft

    Science.gov (United States)

    Wiegmann, Bruce M.

    2017-01-01

    , 3) Controllability of the space-craft via a voltage bias to steer itself through the solar system to destinations of discovery. These activities once demonstrated analytically, will require a technology demonstration mission (TDM) around the year2020 to demonstrate that all systems work together seamlessly before a Heliophysics Electrostatic Rapid Transit System (HERTS) mission could be initiated. A notional TDM spacecraft that meets the requirements of such a mission will be showcased in this paper.

  9. The plant design analyser and its applications

    International Nuclear Information System (INIS)

    Whitmarsh-Everiss, M.J.

    1992-01-01

    Consideration is given to the history of computational methods for the non-linear dynamic analysis of plant behaviour. This is traced from analogue to hybrid computers. When these were phased out simulation languages were used in the batch mode and the interactive computational capabilities were lost. These have subsequently been recovered using mainframe computing architecture in the context of small models using the Prototype Plant Design Analyser. Given the development of parallel processing architectures, the restriction on model size can be lifted. This capability and the use of advanced Work Stations and graphics software has enabled an advanced interactive design environment to be developed. This system is generic and can be used, with suitable graphics development, to study the dynamics and control behaviour of any plant or system for minimum cost. Examples of past and possible future uses are identified. (author)

  10. Development of inelastic design method for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Takahashi, Yukio; Take, Kohji; Kaguchi, Hitoshi; Fukuda, Yoshio; Uno, Tetsuro.

    1991-01-01

    Effective utilization of inelastic analysis in structural design assessment is expected to play an important role for avoiding too conservative design of liquid metal reactor plants. Studies have been conducted by the authors to develop a guideline for application of detailed inelastic analysis in design assessment. Both fundamental material characteristics tests and structural failure tests were conducted. Fundamental investigations were made on inelastic analysis method and creep-fatigue life prediction method based on the results of material characteristics tests. It was demonstrated through structural failure tests that the design method constructed based on these fundamental investigations can predict failure lives in structures subjected to cyclic thermal loadings with sufficient accuracy. (author)

  11. Quality planning for major plant design modifications

    International Nuclear Information System (INIS)

    Dulee, R.J.

    1988-01-01

    This paper reviews the approach and activities undertaken by Public Service Electric and Gas Company's (PSE and G's) nuclear quality assurance (QA) department to support major plant design modifications conducted during refueling outages at Salem Generating Station. It includes the planning and implementation of quality plans developed to provide both QA and quality control (QC) coverage of modification performed by contracted service organizations

  12. New student-designed research and demonstration drop tower

    Science.gov (United States)

    Bell, Donald; Weislogel, Mark

    A new drop tower has been designed and constructed at Portland State University. The ap-proach incorporates innovative features to increase throughput and microgravity quality in a highly public facility. Push button operation with full wireless CCTV coverage and passive magnetic deceleration provides quiet, safe operation from a single control station with low re-cycle time. A two-stage coaxial release mechanism decouples the payload from the drag shield to minimize disturbances to the experiment during release. This is especially important for fluids experiments that are highly sensitive to initial conditions. Performance of the new tower is presented including release, free fall, and deceleration accelerometer data. The two second tower is used for research and educational outreach. The research efforts focus on capillary flows and phenomena relevant to spacecraft fluid systems. The outreach efforts utilize partnerships with local primary, secondary and post-secondary institutions to promote the fields of science, technology, engineering and mathematics.

  13. Feasibility design study. Land-based OTEC plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Brewer, J. H.; Minor, J.; Jacobs, R.

    1979-01-01

    The purpose of this study has been to determine the feasibility of installing 10 MWe (MegaWatt-electric) and 40 MWe land-based OTEC demonstration power plants at two specific sites: Keahole Point on the western shore of the island of Hawaii; and Punta Tuna, on the southeast coast of the main island of Puerto Rico. In addition, the study has included development of design parameters, schedules and budgets for the design, construction and operation of these plants. Seawater systems (intake and discharge pipes) were to be sized so that flow losses were equivalent to those expected with a platform-based OTEC power plant. The power module (components and general arrangement was established based on the TRW design. Results are presented in detail. (WHK)

  14. Optimal design of regional wastewater pipelines and treatment plant systems.

    Science.gov (United States)

    Brand, Noam; Ostfeld, Avi

    2011-01-01

    This manuscript describes the application of a genetic algorithm model for the optimal design of regional wastewater systems comprised of transmission gravitational and pumping sewer pipelines, decentralized treatment plants, and end users of reclaimed wastewater. The algorithm seeks the diameter size of the designed pipelines and their flow distribution simultaneously, the number of treatment plants and their size and location, the pump power, and the required excavation work. The model capabilities are demonstrated through a simplified example application using base runs and sensitivity analyses. Scaling of the proposed methodology to real life wastewater collection and treatment plants design problems needs further testing and developments. The model is coded in MATLAB using the GATOOL toolbox and is available from the authors.

  15. Integrated LED/Imaging Illumination Panels Demonstrated within a Small Plant Growth Chamber Project

    Data.gov (United States)

    National Aeronautics and Space Administration — LED light sources are ideal for plant growth systems. However, commercially available multi-color LED illumination panels are designed and manufactured to produce a...

  16. Investigation of human system interface design in nuclear power plant

    International Nuclear Information System (INIS)

    Feng Yan; Zhang Yunbo; Wang Zhongqiu

    2012-01-01

    The paper introduces the importance of HFE in designing nuclear power plant, and introduces briefly the content and scope of HFE, discusses human system interface design of new built nuclear power plants. This paper also describes human system interface design of foreign nuclear power plant, and describes in detail human system interface design of domestic nuclear power plant. (authors)

  17. Role of Fugen HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, Sadamu.

    1982-03-01

    Fugen, a 165 MWe prototype of a heavy water-moderated, boiling light water-cooled reactor, has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work for a 600 MWe demonstration plant has been carried out since 1973. The important systems and components, such as pressure tube assemblies and control rod drive mechanism, are essentially the same as those of Fugen. However, some modification is made owing to the experience obtained in Fugen and LWrs. In the HWR Fugen, plutonium and uranium are effectively used, and plutonium makes the coolant void reactivity more negative, which results in the increase of the stability and safety of the reactor. On August 4, 1981, the ad hoc committee submitted the final report to the Japanese Atomic Energy Commission, in which the construction of a 600 MWe demonstration plant was recommended. As for the research and development on reactor safety, coolant leak detectors, the performance of ECCS, and safety design codes are enumerated. Since 1965, mixed oxide fuel has been developed, and 168 fuel assemblies were loaded in Fugen, but failure did not occur. (Kako, I.)

  18. System 80+TM standard plant: Design and operations overview

    International Nuclear Information System (INIS)

    Matzie, R.A.; Ritterbusch, S.E.

    1999-01-01

    The System 80+ Standard Plant Design is a 1400 MWe evolutionary Advanced Light Water Reactor (ALWR), designed to meet the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) and the demands of the international market for nuclear power plants which are not only safer but also more economical to maintain and operate. ABB Combustion Engineering Nuclear Power used a defense-in-depth process that (1) adds design margin to basic components to improve performance during normal operation and to decrease the likelihood of an unanticipated transient or an accident, (2) improves the redundancy and diversity of safety systems in order to mitigate design basis accidents and prevent severe accidents, and (3) improves severe accident mitigation capability. This paper describes the most important improved systems and components with emphasis on severe accident prevention and mitigation capability. The improved design features were implemented in an evolutionary manner using proven components. This approach ensures that the plant operates safely and economically, as demonstrated by operating plants in the US and the Republic of Korea. Detailed studies, summarized in this paper, have shown that the System 80+ plant availability is expected to exceed the ALWR requirement of 87% and that the annual operations and maintenance costs are expected to be reduced by $14 million. (author)

  19. Structural modules in AP1000 plant design

    International Nuclear Information System (INIS)

    Prasad, N.; Tunon-Sanjur, L.

    2007-01-01

    Structural modules are extensively used in AP1000 plant design. The shop manufacturing of modules components improves the quality and reliability of plant structures. The application of modules has a positive impact on construction schedules, and results in substantial savings in the construction cost. This paper describes various types of structural modules used for AP1000 plant structures. CA structural wall modules are steel plate modules with concrete placed, on or within the module, after module installation. The layout and design of the largest CA wall modules, CA01 and CA20, is described in detail. General discussion of structural floor modules, such as the composite and finned floors, is also included. Steel form CB modules (liners) consist of plate reinforced with angle stiffeners and tee sections. The angles and the tee sections are on the concrete side of the plate. Design of CB20 has been included as an example of CB type modules. Design codes and structural concepts related to module designs are discussed. (authors)

  20. Design of 90-mgd wastewater reclamation plant

    International Nuclear Information System (INIS)

    Cain, C.B.; Kluesener, J.W.; Lazarus, E.

    1981-01-01

    Ninety MGD of municipal secondary effluent is to be reclaimed for use as cooling water in the Palo Verde Nuclear Generating Station in Arizona. The water reclamation plant design criteria are presented for nitrification, lime-soda softening, filtration, and sludge processing. In-pipe storage is used for flow equalization. Forced air down-draft trickling filters with computer-controlled recycle are used for nitrification. Effluent recirculation is used to maintain constant sludge blanket dept in the upflow solids contact clarifier softening units. An influent flow-splitting, gravity-backwash, dual-media filter is used for final suspended solids removal. Rheological data was used in the hydraulic design of sludge piping. Computerized automatic flushing also aids in preventing sludge pumping problems. Reclamation plant design and construction are integrated in one Engineering-Procurement-Construction project conducted by a single company. 5 refs

  1. Integrated design of SIGMA uranium enrichment plants

    International Nuclear Information System (INIS)

    Rivarola, Martin E.; Brasnarof, Daniel O.

    1999-01-01

    In the present work, we describe a preliminary analysis of the design feedbacks in a Uranium Enrichment Plant, using the SIGMA concept. Starting from the result of this analysis, a computer code has been generated, which allows finding the optimal configurations of plants, for a fixed production rate. The computer code developed includes the model of the Thermohydraulic loop of a SIGMA module. The model contains numerical calculations of the main components of the circuit. During the calculations, the main components are dimensioned, for a posterior cost compute. The program also makes an estimation of the enrichment gain of the porous membrane, for each separation stage. Once the dimensions of the main components are known, using the enrichment cascade calculation, the capital and operation costs of the plant could be determined. At this point it is simple to calculate a leveled cost of the Separative Work Unit (SWU). A numerical optimizer is also included in the program. This optimizer finds the optimal cascade configuration, for a given set of design parameters. The whole-integrated program permits to investigate in detail the feedback in the component design. Therefore, the sensibility of the more relevant parameters can be computed, with respect of the economical variables of the plant. (author)

  2. Real Time Demonstration Project XRF Performance Evaluation Report for Paducah Gaseous Diffusion Plant AOC 492

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Robert L [Argonne National Laboratory

    2008-04-03

    This activity was undertaken to demonstrate the applicability of market-available XRF instruments to quantify metal concentrations relative to background and risk-based action and no action levels in Paducah Gaseous Diffusion Plant (PGDP) soils. As such, the analysis below demonstrates the capabilities of the instruments relative to soil characterization applications at the PGDP.

  3. The Hidden Habit of the Entomopathogenic Fungus Beauveria bassiana: First Demonstration of Vertical Plant Transmission

    Science.gov (United States)

    Quesada-Moraga, Enrique; López-Díaz, Cristina; Landa, Blanca Beatriz

    2014-01-01

    Beauveria bassiana strain 04/01-Tip, obtained from a larva of the opium poppy stem gall wasp Iraella luteipes (Hymenoptera; Cynipidae), endophytically colonizes opium poppy (Papaver somniferum L.) plants and protects them against this pest. The goal of this study was to monitor the dynamics of endophytic colonization of opium poppy by B. bassiana after the fungus was applied to the seed and to ascertain whether the fungus is transmitted vertically via seeds. Using a species-specific nested PCR protocol and DNA extracted from surface-sterilised leaf pieces or seeds of B. bassiana-inoculated opium poppy plants, the fungus was detected within the plant beginning at the growth stage of rosette building and them throughout the entire plant growth cycle (about 120–140 days after sowing). The fungus was also detected in seeds from 50% of the capsules sampled. Seeds that showed positive amplification for B. bassiana were planted in sterile soil and the endophyte was again detected in more than 42% of the plants sampled during all plant growth stages. Beauveria bassiana was transmitted to seeds in 25% of the plants from the second generation that formed a mature capsule. These results demonstrate for the first time the vertical transmission of an entomopathogenic fungus from endophytically colonised maternal plants. This information is crucial to better understand the ecological role of entomopathogenic fungi as plant endophytes and may allow development of a sustainable and cost effective strategy for I. luteipes management in P. somniferum. PMID:24551242

  4. The hidden habit of the entomopathogenic fungus Beauveria bassiana: first demonstration of vertical plant transmission.

    Science.gov (United States)

    Quesada-Moraga, Enrique; López-Díaz, Cristina; Landa, Blanca Beatriz

    2014-01-01

    Beauveria bassiana strain 04/01-Tip, obtained from a larva of the opium poppy stem gall wasp Iraella luteipes (Hymenoptera; Cynipidae), endophytically colonizes opium poppy (Papaver somniferum L.) plants and protects them against this pest. The goal of this study was to monitor the dynamics of endophytic colonization of opium poppy by B. bassiana after the fungus was applied to the seed and to ascertain whether the fungus is transmitted vertically via seeds. Using a species-specific nested PCR protocol and DNA extracted from surface-sterilised leaf pieces or seeds of B. bassiana-inoculated opium poppy plants, the fungus was detected within the plant beginning at the growth stage of rosette building and them throughout the entire plant growth cycle (about 120-140 days after sowing). The fungus was also detected in seeds from 50% of the capsules sampled. Seeds that showed positive amplification for B. bassiana were planted in sterile soil and the endophyte was again detected in more than 42% of the plants sampled during all plant growth stages. Beauveria bassiana was transmitted to seeds in 25% of the plants from the second generation that formed a mature capsule. These results demonstrate for the first time the vertical transmission of an entomopathogenic fungus from endophytically colonised maternal plants. This information is crucial to better understand the ecological role of entomopathogenic fungi as plant endophytes and may allow development of a sustainable and cost effective strategy for I. luteipes management in P. somniferum.

  5. Advanced Neutron Sources: Plant Design Requirements

    International Nuclear Information System (INIS)

    1990-07-01

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW th , heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS

  6. Heat exchanger design for desalination plants

    International Nuclear Information System (INIS)

    1979-03-01

    The Office of Saline Water (OSW) accomplished a very large amount of significant work related to the design and performance of large heat exchanger bundles and enhanced heat transfer surfaces. This work was undertaken to provide basic technical and economic data for the design of distillation plants for the desalination of seawater, and should be of value to other industrial applications as well. The OSW work covers almost every aspect of heat exchanger design, and ranges academic research to data gathering on commercial desalting plants. Exchanger design configurations include multistage flash plant condensers, vertical tube falling film and upflow evaporators, and horizontal tube spray film evaporators. Unfortunately, the data is scattered through a large number of reports of which many are concerned primarily with factors other than heat transfer, and the quality of reporting and the quality of the data are far from consistent. This report catalogues and organizes the heat exchanger data developed by the OSW. Some analysis as to the validity of the data is made and ranges of performance that can be expected are given. Emphasis is placed on the vertical tube, falling film evaporators. A thorough analysis of the large literature file that was surveyed was not possible. No analysis was made of the quality of original data, but apparent data discrepancies are pointed out where such discrepancies happen to be found

  7. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant Conceptual Design Engineering Report (CDER)

    Science.gov (United States)

    1981-01-01

    The reference conceptual design of the magnetohydrodynamic (MHD) Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD, is summarized. Main elements of the design, systems, and plant facilities are illustrated. System design descriptions are included for closed cycle cooling water, industrial gas systems, fuel oil, boiler flue gas, coal management, seed management, slag management, plant industrial waste, fire service water, oxidant supply, MHD power ventilating

  8. Risk informed life cycle plant design

    International Nuclear Information System (INIS)

    Hill, Ralph S. III; Nutt, Mark M.

    2003-01-01

    Many facility life cycle activities including design, construction, fabrication, inspection and maintenance are evolving from a deterministic to a risk-informed basis. The risk informed approach uses probabilistic methods to evaluate the contribution of individual system components to total system performance. Total system performance considers both safety and cost considerations including system failure, reliability, and availability. By necessity, a risk-informed approach considers both the component's life cycle and the life cycle of the system. In the nuclear industry, risk-informed approaches, namely probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA), have become a standard tool used to evaluate the safety of nuclear power plants. Recent studies pertaining to advanced reactor development have indicated that these new power plants must provide enhanced safety over existing nuclear facilities and be cost-competitive with other energy sources. Risk-informed approaches, beyond traditional PRA, offer the opportunity to optimize design while considering the total life cycle of the plant in order to realize these goals. The use of risk-informed design approaches in the nuclear industry is only beginning, with recent promulgation of risk-informed regulations and proposals for risk-informed codes. This paper briefly summarizes the current state of affairs regarding the use of risk-informed approaches in design. Key points to fully realize the benefit of applying a risk-informed approach to nuclear power plant design are then presented. These points are equally applicable to non-nuclear facilities where optimization for cost competitiveness and/or safety is desired. (author)

  9. Power conversion and quality of the Santa Clara 2 MW direct carbonate fuel cell demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Skok, A.J. [Fuel Cell Engineering Corp., Danbury, CT (United States); Abueg, R.Z. [Basic Measuring Instruments, Santa Clara, CA (United States); Schwartz, P. [Fluor Daniel, Inc., Irvine, CA (United States)] [and others

    1996-12-31

    The Santa Clara Demonstration Project (SCDP) is the first application of a commercial-scale carbonate fuel cell power plant on a US electric utility system. It is also the largest fuel cell power plant ever operated in the United States. The 2MW plant, located in Santa Clara, California, utilizes carbonate fuel cell technology developed by Energy Research Corporation (ERC) of Danbury, Connecticut. The ultimate goal of a fuel cell power plant is to deliver usable power into an electrical distribution system. The power conversion sub-system does this for the Santa Clara Demonstration Plant. A description of this sub-system and its capabilities follows. The sub-system has demonstrated the capability to deliver real power, reactive power and to absorb reactive power on a utility grid. The sub-system can be operated in the same manner as a conventional rotating generator except with enhanced capabilities for reactive power. Measurements demonstrated the power quality from the plant in various operating modes was high quality utility grade power.

  10. Role of Fugen-HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, S.

    1982-01-01

    Fugen, a 165 MWe prototype of a heavy water moderated boiling light water cooled reactor; has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work of the 600 MWe demonstration plant has been carried out since 1973. Important system and components, such as pressure tube assemblies, control rod drive mechanism, etc., are essentially the same as those of Fugen. Some modifications, however, are made especially from the stand point of experiences In the Fugen-HWR, plutonium and uranium would be effectively used; and plutonium could make the coolant void reactivity more negative which would give good results in increasing the reactor stability and safety. On the other hand, nuclear power plants are mainly consisted of LWRs in Japan. Considering the above situations, the Fugen-HWR, coupled with LWRs, is now considered in Japan to contribute to our energy security by using plutonium and depleted uranium extracted from spent fuels of LWRs: thereby reducing the demands On August 4, 1981, the ad hoc committee on the 600 MWe demonstration Fugen-HWR submitted the final report to the Japan AEC, after having had discussions and evaluations. In the report, the ad hoc committee recommended to build the 600 MWE demonstration plant with appropriate supports of the Government. The Japan AEC will be expected to make her decision on the program in the near future. As for the reactor safety R and C, development has been stressed on coolant leak detectors and ECCS performances or Since 1965, many development works have been done for mixed oxide fuel assemblies, both for establishment of the fabrication technology and for clarification of irradiation performances. 196 mixed oxide fuel assemblies have been manufactured for Fugen. 168 of them were loaded and 92 were withdrawn. No fuel has been failured yet. (author)

  11. DU-AGG pilot plant design study

    International Nuclear Information System (INIS)

    Lessing, P.A.; Gillman, H.

    1996-07-01

    The Idaho National Engineering Laboratory (INEL) is developing new methods to produce high-density aggregate (artificial rock) primarily consisting of depleted uranium oxide. The objective is to develop a low-cost method whereby uranium oxide powder (UO[sub 2], U[sub 3]O[sub ]8, or UO[sub 3]) can be processed to produce high-density aggregate pieces (DU-AGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive materials. A commercial company, G-M Systems, conducted a design study for a manufacturing pilot plant to process DU-AGG. The results of that study are included and summarized in this report. Also explained are design considerations, equipment capacities, the equipment list, system operation, layout of equipment in the plant, cost estimates, and the proposed plan and schedule

  12. Fiscal 1995 achievement report. Development of entrained bed coal gasification power plant (Part 5 - Surveys and studies of demonstration plant); 1995 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 5. Jissho plant ni kansuru chosa kenkyu hen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    Surveys and studies were conducted concerning a demonstration plant for establishing the technology of integrated coal gasification combined cycle, and the fiscal 1995 results are compiled. In this fiscal year, a demonstration plant conceptual design was prepared for assuring smooth transition from a pilot plant to a commercial plant. The design followed the system employed at the Nakoso pilot plant for its gasification power generation. It was decided that the gasification furnace be of the air-blown (oxygen enriched) 2-stage entrained bed type, that the desulfurization system be of the dry type 2-stage fluidized bed type, the dedusting system be of the dry type granular bed type (moving bed type), that the combined cycle power facility be derived from the commercialized gas turbine, and that the cycle of the steam system agree with the integrated coal gasification combined cycle system now under discussion. Studies were made, which covered heat efficiency (generating end/sending end), heat/matter balance, process flow, gas turbine/steam system optimization, comparison in performance with a pilot plant with its dimensions increased, estimation of the performance of each of the facilities, estimation of the construction cost, calculation of the generation cost, environmental friendliness, operating characteristics, acceptable coal types, and the like. (NEDO)

  13. Plant Design for the Production of DUAGG

    International Nuclear Information System (INIS)

    Ferrada, J.J.

    2003-01-01

    The cost of producing DUAGG is an important consideration for any interested private firm in determining whether DUCRETE is economically viable as a material of construction in next-generation spent nuclear fuel casks. This study analyzed this project as if it was a stand-alone project. The capital cost includes engineering design, equipment costs and installation, start up, and management; the study is not intended to be a life-cycle cost analysis. The costs estimated by this study are shown in Table ES.1, and the conclusions of this study are listed in Table ES.2. The development of DUAGG and DUCRETE is a major thrust of the Depleted Uranium Uses Research and Development Project. An obvious use of depleted uranium is as a shielding material (e.g., DUCRETE). DUCRETE is made by replacing the conventional stone aggregate in concrete with DUAGG. One objective of this project is to bring the development of DUCRETE to a point at which a demonstrated basis exists for its commercial deployment. The estimation of the costs to manufacture DUAGG is an important part of this effort. Paul Lessing and William Quapp developed DUAGG and DUCRETE as part of an Idaho National Engineering and Environmental Laboratory (INEEL) program to find beneficial uses for depleted uranium (DU). Subsequently, this technology was licensed to Teton Technologies, Inc. The DUAGG process mixes DUO 2 with sintering materials and additives to form pressed briquettes. These briquettes are sintered at 1300 C, and the very dense sintered briquettes are then crushed and classified into gap-graded size fractions. The graded DUAGG is then ready to be used to make high-strength heavy DUCRETE. The DUCRETE shielding will be placed into an annular steel cask-shell mold, which has internal steel reinforcing bars. The objectives of this study are to (1) use previous DUAGG process developments to design a plant that will produce DUAGG at a baseline rate, (2) determine the size of the equipment required to meet the

  14. Exergy analysis of an IGCC design configuration for Plant Wansley

    International Nuclear Information System (INIS)

    Tsatsaronis, G.; Tawfik, T.; Lin, L.; Gallaspy, D.T.

    1989-01-01

    An integrated gasification-combined-cycle power plant design was developed for Georgia Power Company's Plant Wansley. This paper discusses the plant configuration and presents the most important results obtained from a detailed exergy analysis of the plant design. This analysis will be completed in a subsequent paper through an exergoeconomic analysis to identify design improvements for reducing the electricity cost

  15. De-novo design of antimicrobial peptides for plant protection.

    Directory of Open Access Journals (Sweden)

    Benjamin Zeitler

    Full Text Available This work describes the de-novo design of peptides that inhibit a broad range of plant pathogens. Four structurally different groups of peptides were developed that differ in size and position of their charged and hydrophobic clusters and were assayed for their ability to inhibit bacterial growth and fungal spore germination. Several peptides are highly active at concentrations between 0,1 and 1 µg/ml against plant pathogenic bacteria, such as Pseudomonas syringae, Pectobacterium carotovorum, and Xanthomonas vesicatoria. Importantly, no hemolytic activity could be detected for these peptides at concentrations up to 200 µg/ml. Moreover, the peptides are also active after spraying on the plant surface demonstrating a possible way of application. In sum, our designed peptides represent new antimicrobial agents and with the increasing demand for antimicrobial compounds for production of "healthy" food, these peptides might serve as templates for novel antibacterial and antifungal agents.

  16. Regulatory issues resolved through design certification on the System 80+trademark standard plant design

    International Nuclear Information System (INIS)

    Ritterbusch, S.E.; Brinkman, C.B.

    1996-01-01

    The US Nuclear Regulatory Commission (NRC) has completed its review of the System 80+trademark Standard Plant Design, approving advanced design features and closing severe accident licensing issues. Final Design Approval was granted in July 1994. The NRC review was extensive, requiring written responses to over 4,950 questions and formal printing of over 50,000 Safety Analysis Report pages. New safety issues never before addressed in a regulatory atmosphere had to be resolved with detailed analysis and evaluation of design features. the System 80+ review demonstrated that regulatory issues can be firmly resolved only through presentation of a detailed design and completion of a comprehensive regulatory review

  17. The IPRP (Integrated Pyrolysis Regenerated Plant) technology: From concept to demonstration

    International Nuclear Information System (INIS)

    D’Alessandro, Bruno; D’Amico, Michele; Desideri, Umberto; Fantozzi, Francesco

    2013-01-01

    Highlights: ► IPRP technology development for distributed conversion of biomass and wastes. ► IPRP demonstrative unit combines a rotary kiln pyrolyzer to a 80 kWe microturbine. ► Main performances and critical issues are pointed out for different residual fuels. -- Abstract: The concept of integrated pyrolysis regenerated plant (IPRP) is based on a Gas Turbine (GT) fuelled by pyrogas produced in a rotary kiln slow pyrolysis reactor, where waste heat from GT is used to sustain the pyrolysis process. The IPRP plant provides a unique solution for microscale (below 250 kW) power plants, opening a new and competitive possibility for distributed biomass or wastes to energy conversion systems. The paper summarizes the state of art of the IPRP technology, from preliminary numerical simulation to pilot plant facility, including some new available data on pyrolysis gas from laboratory and pilot plants.

  18. Plant diversity to support humans in a CELSS ground based demonstrator

    Science.gov (United States)

    Howe, J. M.; Hoff, J. E.

    1981-01-01

    A controlled ecological life support system (CELSS) for human habitation in preparation for future long duration space flights is considered. The success of such a system depends upon the feasibility of revitalization of food resources and the human nutritional needs which are to be met by these food resources. Edible higher plants are prime candidates for the photoautotrophic components of this system if nutritionally adequate diets can be derived from these plant sources to support humans. Human nutritional requirements information based on current knowledge are developed for inhabitants envisioned in the CELSS ground based demonstrator. Groups of plant products that can provide the nutrients are identified.

  19. Conceptual design of small-sized HTGR system (4). Plant design and technical feasibility

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing L.; Sumita, Junya; Nomoto, Yasunobu; Tazawa, Yujiro; Noguchi, Hiroki; Imai, Yoshiyuki; Tachibana, Yukio

    2013-09-01

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine, to deploy in developing countries in the 2020s. HTR50S was designed for steam supply and electricity generation by the steam turbine with the reactor outlet temperature of 750degC as a reference plant configuration. On the other hand, the intermediate heat exchanger (IHX) will be installed in the primary loop to demonstrate the electricity generation by the helium gas turbine and hydrogen production by thermochemical water splitting by utilizing the secondary helium loop with the reactor outlet temperature of 900degC as a future plant configuration. The plant design of HTR50S for the steam supply and electricity generation was performed based on the plant specification and the requirements for each system taking into account for the increase of the reactor outlet coolant temperature from 750degC to 900degC and the installation of IHX. The technical feasibility of HTR50S was confirmed because the designed systems (i.e., reactor internal components, reactor pressure vessel, vessel cooling system, shutdown cooling system, steam generator (SG), gas circulator, SG isolation and drainage system, reactor containment vessel, steam turbine and heat supply system) satisfies the design requirements. The conceptual plant layout was also determined. This paper provides the summary of the plan design and technical feasibility of HTR50S. (author)

  20. St. Louis demonstration final report: refuse processing plant equipment, facilities, and environmental evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Fiscus, D.E.; Gorman, P.G.; Schrag, M.P.; Shannon, L.J.

    1977-09-01

    The results are presented of processing plant evaluations of the St. Louis-Union Electric Refuse Fuel Project, including equipment and facilities as well as assessment of environmental emissions at both the processing and the power plants. Data on plant material flows and operating parameters, plant operating costs, characteristics of plant material flows, and emissions from various processing operations were obtained during a testing program encompassing 53 calendar weeks. Refuse derived fuel (RDF) is the major product (80.6% by weight) of the refuse processing plant, the other being ferrous metal scrap, a marketable by-product. Average operating costs for the entire evaluation period were $8.26/Mg ($7.49/ton). The average overall processing rate for the period was 168 Mg/8-h day (185.5 tons/8-h day) at 31.0 Mg/h (34.2 tons/h). Future plants using an air classification system of the type used at the St. Louis demonstration plant will need an emissions control device for particulates from the large de-entrainment cyclone. Also in the air exhaust from the cyclone were total counts of bacteria and viruses several times higher than those of suburban ambient air. No water effluent or noise exposure problems were encountered, although landfill leachate mixed with ground water could result in contamination, given low dilution rates.

  1. Project designing of Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Krychtalek, Z.; Linek, V.

    1989-01-01

    The geological and seismic parameters are listed of the Temelin nuclear power plant. The division of the site in building zones is described. The main zones consist of the power generation unit zone with the related auxiliary buildings of hot plants and of the auxiliary buildings of the nonactive part with industrial buildings. The important buildings are interconnected with communication and technology bridges. Cooling towers and spray pools and the entrance area are part of the urbanistic design. The architectonic design of the buildings uses standard building elements and materials. The design of the buildings is based on the requirements on their function and on structural load and on the demands of maximal utilization of the type of the reinforced concrete prefab structure system. The structure is made of concrete or steel cells. The project design is based on Soviet projects. The layout is shown of the main power generation units and a section is presented of a 1,000 MW unit. (J.B.). 2 figs

  2. Advanced Grid-Friendly Controls Demonstration Project for Utility-Scale PV Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Gevorgian, Vahan; O' Neill, Barbara

    2016-01-21

    A typical photovoltaic (PV) power plant consists of multiple power electronic inverters and can contribute to grid stability and reliability through sophisticated 'grid-friendly' controls. The availability and dissemination of actual test data showing the viability of advanced utility-scale PV controls among all industry stakeholders can leverage PV's value from being simply an energy resource to providing additional ancillary services that range from variability smoothing and frequency regulation to power quality. Strategically partnering with a selected utility and/or PV power plant operator is a key condition for a successful demonstration project. The U.S. Department of Energy's (DOE's) Solar Energy Technologies Office selected the National Renewable Energy Laboratory (NREL) to be a principal investigator in a two-year project with goals to (1) identify a potential partner(s), (2) develop a detailed scope of work and test plan for a field project to demonstrate the gird-friendly capabilities of utility-scale PV power plants, (3) facilitate conducting actual demonstration tests, and (4) disseminate test results among industry stakeholders via a joint NREL/DOE publication and participation in relevant technical conferences. The project implementation took place in FY 2014 and FY 2015. In FY14, NREL established collaborations with AES and First Solar Electric, LLC, to conduct demonstration testing on their utility-scale PV power plants in Puerto Rico and Texas, respectively, and developed test plans for each partner. Both Puerto Rico Electric Power Authority and the Electric Reliability Council of Texas expressed interest in this project because of the importance of such advanced controls for the reliable operation of their power systems under high penetration levels of variable renewable generation. During FY15, testing was completed on both plants, and a large amount of test data was produced and analyzed that demonstrates the ability of

  3. ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

    Science.gov (United States)

    Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

    2014-10-01

    The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

  4. Conceptual designs parameters for MURR LEU U-Mo fuel conversion design demonstration experiment. Revision 1

    International Nuclear Information System (INIS)

    Stillman, J.; Feldman, E.; Stevens, J.

    2013-01-01

    The design parameters for the conceptual design of a fuel assembly containing U-10Mo fuel foils with low-enriched uranium (LEU) for the University of Missouri Research Reactor (MURR) are described. The Design Demonstration Experiment (MURR-DDE) will use a prototypic MURR-LEU element manufactured according to the parameters specified here. Also provided are calculated performance parameters for the LEU element in the MURR, and a set of goals for the MURR-DDE related to those parameters. The conversion objectives are to develop a fuel element design that will ensure safe reactor operations, as well as maintaining existing performance. The element was designed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. A set of manufacturing assumptions were provided by the Fuel Development (FD) and Fuel Fabrication Capability (FFC) pillars of the GTRI Reduced Enrichment for Research and Test Reactors (RERTR) program to reliably manufacture the fuel plates. The proposed LEU fuel element has an overall design and exterior dimensions that are similar to those of the current highly-enriched uranium (HEU) fuel elements. There are 23 fuel plates in the LEU design. The overall thickness of each plate is 44 mil, except for the exterior plate that is furthest from the center flux trap (plate 23), which is 49 mil thick. The proposed LEU fuel plates have U-10Mo monolithic fuel foils with a 235U enrichment of 19.75% varying from 9 mil to 20 mil thick, and clad with Al-6061 aluminum. A thin layer of zirconium exists between the fuel foils and the aluminum as a diffusion barrier. The thinnest nominal combined zirconium and aluminum clad thickness on each side of the fuel plates is 12 mil. The LEU U-10Mo monolithic fuel is not yet qualified as driver fuel in research reactors, but is under intense development under the auspices of the GTRI FD and FFC programs.

  5. Integrated plant information technology design support functionality

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Kim, Dae Jin; Barber, P. W.; Goland, D.

    1996-06-01

    This technical report was written as a result of Integrated Plant Information System (IPIS) feasibility study on CANDU 9 project which had been carried out from January, 1994 to March, 1994 at AECL (Atomic Energy Canada Limited) in Canada. From 1987, AECL had done endeavour to change engineering work process from paper based work process to computer based work process through CANDU 3 project. Even though AECL had a lot of good results form computerizing the Process Engineering, Instrumentation Control and Electrical Engineering, Mechanical Engineering, Computer Aided Design and Drafting, and Document Management System, but there remains the problem of information isolation and integration. On this feasibility study, IPIS design support functionality guideline was suggested by evaluating current AECL CAE tools, analyzing computer aided engineering task and work flow, investigating request for implementing integrated computer aided engineering and describing Korean request for future CANDU design including CANDU 9. 6 figs. (Author)

  6. Integrated plant information technology design support functionality

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Seung; Kim, Dae Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Barber, P W; Goland, D [Atomic Energy Canada Ltd., (Canada)

    1996-06-01

    This technical report was written as a result of Integrated Plant Information System (IPIS) feasibility study on CANDU 9 project which had been carried out from January, 1994 to March, 1994 at AECL (Atomic Energy Canada Limited) in Canada. From 1987, AECL had done endeavour to change engineering work process from paper based work process to computer based work process through CANDU 3 project. Even though AECL had a lot of good results form computerizing the Process Engineering, Instrumentation Control and Electrical Engineering, Mechanical Engineering, Computer Aided Design and Drafting, and Document Management System, but there remains the problem of information isolation and integration. On this feasibility study, IPIS design support functionality guideline was suggested by evaluating current AECL CAE tools, analyzing computer aided engineering task and work flow, investigating request for implementing integrated computer aided engineering and describing Korean request for future CANDU design including CANDU 9. 6 figs. (Author).

  7. Economic comparison of fusion power plant designs

    International Nuclear Information System (INIS)

    O'Neill, J.E.

    1986-01-01

    Over the past 10 yr, a number of studies have been developed for fusion power plants of various types (tokamaks, mirrors, etc.) complete with figures of merit such as cost estimates and estimates of the cost of generating electricity (COE). Each of these designs involves unresolved physics and engineering problems which, it is assumed, will eventually be worked out. Because of such uncertainties the figures of merit associated with such designs are not to be compared as absolute measures of worth but as relative indicators of progress within a given concept type. As part of Grumman's involvement in fusion energy development, an effort has been undertaken to compare economic indicators from the referenced studies in order to determine the cost trend in recent reactor design activities

  8. Demonstration project: Oxy-fuel combustion at Callide-A plant

    Energy Technology Data Exchange (ETDEWEB)

    Makino, Keiji; Misawa, Nobuhiro; Kiga, Takashi; Spero, Chris

    2007-07-01

    Oxy-fuel combustion is expected to be one of the promising systems on CO2 recovery from pulverized-coal power plant, and enable the CO2 to be captured in a more cost-effective manner compared to other CO2 recover process. An Australia-Japan consortium was established in 2004 specifically for the purpose of conducting a feasibility study on the application of oxy-fuel combustion to an existing pulverized-coal power plant that is Callide-A power plant No.4 unit at 30MWe owned by CS Energy in Australia. One of the important components in this study has been the recent comparative testing of three Australian coals under both oxy-fuel and air combustion conditions using the IHI combustion test facilities. The tests have yielded a number of important outcomes including a good comparison of normal air with oxy-fuel combustion, significant reduction in NOx mass emission rates under oxy-fuel combustion. On the basis of the feasibility study, the project under Australia-Japan consortium is now under way for applying oxy-fuel combustion to an existing plant by way of demonstration. In this project, a demonstration plant of oxy-fuel combustion will be completed by the end of 2008. This project aims at recovering CO2 from an actual power plant for storage. (auth)

  9. Design-Only Conceptual Design Report: Plutonium Immobilization Plant

    International Nuclear Information System (INIS)

    DiSabatino, A.; Loftus, D.

    1999-01-01

    This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

  10. Design report small-scale fuel alcohol plant. Volume 2: Detailed construction information

    Science.gov (United States)

    1980-12-01

    The objectives are to provide potential alcohol producers with a reference design and provide a complete, demonstrated design of a small scale fuel alcohol plant. The plant has the capability for feedstock preparation, cooking, saccharification, fermentation, distillation, by-product dewatering, and process steam generation. An interesting feature is an instrumentation and control system designed to allow the plant to run 24 hours per day with only four hours of operator attention.

  11. RIMAP demonstration project. Pat. 1: Risk based life management of piping system in power plant Heilbronn

    International Nuclear Information System (INIS)

    Bareiss, J.; Puck, P.; Matschecko, B.; Jovanovic, A.; Balos, D.; Perunicic

    2003-01-01

    In the framework of EU project RIMAP [1] a new European Guideline for optimized risk based maintenance and inspection planning of industrial plants (RBLM - Risk Based Life Management) is being developed. The RIMAP project consists of the three clustered projects: development (RTD), demonstration (DEMO) and thematic network (TN). Current work and future, planned work in RIMAP demonstration project on applications of the RIMAP methodology in power plants are presented briefly in the first part of the paper. Also presented in the paper are the results of a preliminary analysis of piping system in power plant Heilbronn using the concept of risk-based monitoring as part of overall concept of risk-based life management. Shortly the following issues are discussed in the paper: identification of critical components, application of a multilevel risk analysis (..from ''screening'' to ''detailed analysis''), determination of PoF - Probability of Failure, determination of COF - Consequence of Failure and optimation of inspection and maintenance plan. (orig.)

  12. Wood gasification demonstration plant in the Schwaebische Alb mountains; Demonstrationsprojekt zur Holzvergasung auf der Schwaebischen Alb

    Energy Technology Data Exchange (ETDEWEB)

    Naab, Peter; Bernhart, Martin [Energieversorgung Filstal GmbH und Co. KG, Goeppingen (Germany)

    2009-12-15

    From 2011, the demonstration plant ''Technologieplattform Bioenergie und Methan (TBM) will produce a hydrogen-rich fuel gas from wood, biomass residues and steam in the intercommunal industrial area ''Gewerbepark Schwaebische Alb'' near Geislingen-Tuerkheim in the German state of Baden-Wuerttemberg. (orig.)

  13. A Miniature Wastewater Cleaning Plant to Demonstrate Primary Treatment in the Classroom

    Science.gov (United States)

    Ne´el, Bastien; Cardoso, Catia; Perret, Didier; Bakker, Eric

    2015-01-01

    A small-scale wastewater cleaning plant is described that includes the key physical pretreatment steps followed by the chemical treatment of mud by flocculation. Water, clay particles, and riverside deposits mimicked odorless wastewater. After a demonstration of the optimization step, the flocculation process was carried out with iron(III)…

  14. Proceedings of the 1978 symposium on instrumentation and control for fossil demonstration plants

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    The 1978 symposium on instrumentation and control for fossil demonstration plants was held at Newport Beach, California, June 19--21, 1978. It was sponsored by Argonne National Laboratory, the U.S. Department of Energy - Fossil Energy, and the Instrument Society of America - Orange County Section. Thirty-nine papers have been entered individually into the data base. (LTN)

  15. New nuclear plant design and licensing process

    International Nuclear Information System (INIS)

    Luangdilok, W.

    1996-01-01

    This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

  16. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

  17. Biorefinery plant design, engineering and process optimisation

    DEFF Research Database (Denmark)

    Holm-Nielsen, Jens Bo; Ehimen, Ehiazesebhor Augustine

    2014-01-01

    Before new biorefinery systems can be implemented, or the modification of existing single product biomass processing units into biorefineries can be carried out, proper planning of the intended biorefinery scheme must be performed initially. This chapter outlines design and synthesis approaches...... applicable for the planning and upgrading of intended biorefinery systems, and includes discussions on the operation of an existing lignocellulosic-based biorefinery platform. Furthermore, technical considerations and tools (i.e., process analytical tools) which could be applied to optimise the operations...... of existing and potential biorefinery plants are elucidated....

  18. Biogas plants: Design, construction and operation

    International Nuclear Information System (INIS)

    2001-01-01

    At the big readiness of waste coming from the agricultural activities are looked for the production of Energy and Payments, the biogas like product of the organic decomposition under anaerobic conditions, their composition and characteristic. The elements that conform the design as the digester, the storage, the load tanks and it discharges and the conduction is described and analyzed. They are given a series of elements to obtain the characteristics of the system possible to place as: planning, calculations, evaluation, execution and operation. Lastly the steps are indicated that should be continued in the construction of the plant including planning for the work

  19. Designs of new plants of high capacity

    International Nuclear Information System (INIS)

    Borges R, Diego A

    1999-01-01

    The Caracas electricity in their desire to lend the best service to the community is doing the necessary projects of generation expansion to fulfill and requirements of demand of next decade in to the metropolitan area. The projects of the new plants of Recifes and El Sitio, have been conceptualized in way of executing the engineering and patterns construction of highest quality and once setting in service, to reach the highest indexes of operative. To reach these goals it is planned to use the most advanced technological designs that are in the market at the moment to world in generation and transmission of power

  20. Spray pond design for nuclear power plants

    International Nuclear Information System (INIS)

    Codell, R.B.; Asce, A.M.

    1986-01-01

    This paper presents a complex methodology for assessing the performance of spray ponds in ultimate heat sink service at nuclear power plants. A spray pond performance model, developed in the companion paper, is used in conjunction with on-site and off-site meteorological data to predict the highest temperature and greatest 30 day water loss which can reasonable be expected to occur during the lifetime of the plant. The performance model for heat and mass transfer is used to develop an efficient phenomenological model used to scan the long-term meteorological records. Refined estimates of temperature or water loss may then be based on more complicated models if necessary. Short-term onsite data are correlated to the long-term off-site data to formulate correction factors for the difference in location. Cumulative distribution functions for temperature and water loss are determined from the long-term meteorological records to predict the occurrence of these quantities which are less severe that the peak. The methodology is demonstrated using data and parameters from the Palo Verde nuclear plant as an example

  1. Preparation of plant and system design description documents

    International Nuclear Information System (INIS)

    1989-01-01

    This standard prescribes the purpose, scope, organization, and content of plant design requirements (PDR) documents and system design descriptions (SDDs), to provide a unified approach to their preparation and use by a project as the principal means to establish the plant design requirements and to establish, describe, and control the individual system designs from conception and throughout the lifetime of the plant. The Electric Power Research Institute's Advanced Light Water Reactor (LWR) Requirements Document should be considered for LWR plants

  2. Demonstration of Essential Reliability Services by a 300-MW Solar Photovoltaic Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Loutan, Clyde; Klauer, Peter; Chowdhury, Sirajul; Hall, Stephen; Morjaria, Mahesh; Chadliev, Vladimir; Milam, Nick; Milan, Christopher; Gevorgian, Vahan

    2017-03-24

    The California Independent System Operator (CAISO), First Solar, and the National Renewable Energy Laboratory (NREL) conducted a demonstration project on a large utility-scale photovoltaic (PV) power plant in California to test its ability to provide essential ancillary services to the electric grid. With increasing shares of solar- and wind-generated energy on the electric grid, traditional generation resources equipped with automatic governor control (AGC) and automatic voltage regulation controls -- specifically, fossil thermal -- are being displaced. The deployment of utility-scale, grid-friendly PV power plants that incorporate advanced capabilities to support grid stability and reliability is essential for the large-scale integration of PV generation into the electric power grid, among other technical requirements. A typical PV power plant consists of multiple power electronic inverters and can contribute to grid stability and reliability through sophisticated 'grid-friendly' controls. In this way, PV power plants can be used to mitigate the impact of variability on the grid, a role typically reserved for conventional generators. In August 2016, testing was completed on First Solar's 300-MW PV power plant, and a large amount of test data was produced and analyzed that demonstrates the ability of PV power plants to use grid-friendly controls to provide essential reliability services. These data showed how the development of advanced power controls can enable PV to become a provider of a wide range of grid services, including spinning reserves, load following, voltage support, ramping, frequency response, variability smoothing, and frequency regulation to power quality. Specifically, the tests conducted included various forms of active power control such as AGC and frequency regulation; droop response; and reactive power, voltage, and power factor controls. This project demonstrated that advanced power electronics and solar generation can be

  3. Aseismic design of Hamaoka Nuclear Power Plant

    International Nuclear Information System (INIS)

    Mizuno, Norihiro

    1975-01-01

    The Hamaoka Nuclear Power Plant of Chubu Electric Power Co. is designed so as to maintain structural safety against an earthquake of 300 gal. For the purpose, a compound-type reactor-housing building is employed, which contains a reactor, operation control and waste disposal facilities. The merits accruing from this scheme are as follows. (1) The shielding walls of the waste disposal facility can be utilized effectively in aseismatic design, leading to the increased rigidity of the building and the uniform distribution of resistance. (2) Due to the large area of the foundation, the load in earthquake can be mitigated, and it resulted in the higher structural stability. Moreover, seismic energy can be dissipated into ground. After the description of the compound building structure, it is explained how the structural resistance and the ground dissipation of seismic energy contribute to potential earthquake resistance. (Mori, K.)

  4. Advanced system demonstration for utilization of biomass as an energy source. Volume I. Scope and design criteria and project summary

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-10-01

    The information in this document is the result of an intensive engineering effort to demonstrate the feasibility of biomass-fueled boilers in cogeneration applications. This design package is based upon a specific site in the State of Maine. However, the design is generic in nature and could serve as a model for other biomass conversion facilities located anywhere biomass is abundant. The project's purpose and summary information are presented: the plant, its concept of operation; and other overall information are described. The capital cost estimate for the plant, and the basis upon which it was obtained are given; a schedule of key milestones and activities required to construct the plant and put it into operation is presented; and the general findings in areas that affect the viability of the project are discussed. The technical design, biomass study, environmental impact, commercialization, and economic factors are addressed. Each major plant area and the equipment and facilities that each includes are discussed in depth. Some overall plant requirements, including noise control, reliability, maintainability, and safety, are detailed. The results of each study relating to alternatives considered for optimizing plant operation parameters and specific system process schemes are briefly presented. All economic factors that affect the feasibility and viability of the biomass project are defined and evaluated.

  5. Safety of Nuclear Power Plants: Design. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  6. Seismic design features of the ACR Nuclear Power Plant

    International Nuclear Information System (INIS)

    Elgohary, M.; Saudy, A.; Aziz, T.

    2003-01-01

    Through their worldwide operating records, CANDU Nuclear Power Plants (NPPs) have repeatedly demonstrated safe, reliable and competitive performance. Currently, there are fourteen CANDU 6 single unit reactors operating or under construction worldwide. Atomic Energy of Canada Limited's (AECL) Advanced CANDU Reactor - the ACR. - is the genesis of a new generation of technologically advanced reactors founded on the CANDU reactor concept. The ACR is the next step in the evolution of the CANDU product line. The ACR products (ACR-700 and ACR-1000) are based on CANDU 6 (700 MWe class) and CANDU 9 (900 MWe class) reactors, therefore continuing AECL's successful approach of offering CANDU plants that appeal to a broad segment of the power generation market. The ACR products are based on the proven CANDU technology and incorporate advanced design technologies. The ACR NPP seismic design complies with Canadian standards that were specifically developed for nuclear seismic design and also with relevant International Atomic Energy Agency (IAEA) Safety Design Standards and Guides. However, since the ACR is also being offered to several markets with many potential sites and different regulatory environments, there is a need to develop a comprehensive approach for the seismic design input parameters. These input parameters are used in the design of the standard ACR product that is suitable for many sites while also maintaining its economic competitiveness. For this purpose, the ACR standard plant is conservatively qualified for a Design Basis Earthquake (DBE) with a peak horizontal ground acceleration of 0.3g for a wide range of soil/rock foundation conditions and Ground Response Spectra (GRS). These input parameters also address some of the current technical issues such as high frequency content and near field effects. In this paper, the ACR seismic design philosophy and seismic design approach for meeting the safety design requirements are reviewed. Also the seismic design

  7. Innovative Offshore Wind Plant Design Study

    Energy Technology Data Exchange (ETDEWEB)

    Hurley, William L. [Glosten Associates, Inc., Seattle, WA (United States); Nordstrom, Charles J. [Glosten Associates, Inc., Seattle, WA (United States); Morrison, Brent J. [Glosten Associates, Inc., Seattle, WA (United States)

    2013-12-18

    Technological advancements in the Glosten PelaStar floating wind turbine system have led to projected cost of energy (COE) reductions from today’s best-in-class offshore wind systems. The PelaStar system is projected to deliver a COE that is 35% lower than that delivered by the current offshore wind plants. Several technology developments have been achieved that directly target significant cost of energy reductions. These include: Application of state-of-the-art steel construction materials and methods, including fatigue-resistant welding techniques and technologies, to reduce hull steel weight; Advancements in synthetic fiber tendon design for the mooring system, which are made possible by laboratory analysis of full-scale sub-rope specimens; Investigations into selected anchor technologies to improve anchor installation methods; Refinement of the installation method, specifically through development of the PelaStar Support Barge design. Together, these technology developments drive down the capital cost and operating cost of offshore wind plants and enable access to superb wind resources in deep water locations. These technology developments also reduce the uncertainty of the PelaStar system costs, which increases confidence in the projected COE reductions.

  8. Design of the vitrification plant for HLLW generated from the Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Vematsu, K.

    1986-01-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) is now designing a vitrification plant. This plant is for the solidification of high-level liquid waste (HLLW) which is generated from the Tokai Reprocessing Plant, and for the demonstration of the vitrification technology. The detailed design of the plant which started in 1982 was completed in 1984. At present the design improvement is being made for the reduction of construction cost and for the licensing which is going to be applied in 1986. The construction will be started in autumn 1987. The plant has a large shielded cell with low flow ventilation, and employs rack-mounted module system and high performance two-armed servomanipulator system to accomplish the fully remote operations and maintenance. The vitrification of HLLW is based on the liquid-fed Joule-heated ceramic melter process. The processing capacity is equivalent to the reprocessing of 0.7 ton of heavy metals per day. The glass production rate is about 9 kg/h, and about 300 kg of glass is poured periodically from the bottom of the melter into a canister. Produced glass is stored under the forced air cooling condition

  9. System design specification Brayton Isotope Power System (BIPS) Flight System (FS), and Ground Demonstration System (GDS)

    International Nuclear Information System (INIS)

    1976-01-01

    The system design specification for ground demonstration, development, and flight qualification of a Brayton Isotope Power System (BIPS) is presented. The requirements for both a BIPS conceptual Flight System (FS) and a Ground Demonstration System (GDS) are defined

  10. Fiscal 1994 achievement report. Development of entrained bed coal gasification power plant (Part 5 - Survey of demonstration plant); 1994 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 5. Jissho plant ni kansuru chosa kenkyu hen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    A survey was conducted of a demonstration plant for developing the technology of integrated coal gasification combined cycle power generation. In fiscal 1994, based on the studies of pilot plant operation conducted up to the preceding fiscal year and on the studies of a demonstration plant (conceptual design of a demonstration plant), systems were examined for still higher performance and economical efficiency. For optimizing the heat source for pulverized coal drying air, an extracted air heat utilizing system was adopted, excellent in performance (approximately 0.5% improvement in absolute value) and in economical efficiency. For reducing the consumption of inert gas for the fluidized bed desulfurization facility, an approximately 30% reduction was found to be possible dependent on gas regeneration conditions in the regeneration tower. For performance improvement with the combined cycle equipment placed under a partial load, the generating-end efficiency was improved approximately 0.2% when only SGC (syngas cooler) was loaded 80% or less and kept at a constant pressure. Studies were also made about how to maintain a constant output with the equipment exposed to air temperature changes. (NEDO)

  11. UNEP Demonstrations of Mercury Emission Reduction at Two Coal-fired Power Plants in Russia

    Directory of Open Access Journals (Sweden)

    Jozewicz W.

    2013-04-01

    Full Text Available The United Nations Environment Programme (UNEP partnership area “Mercury releases from coal combustion” (The UNEP Coal Partnership has initiated demonstrations of mercury air emission reduction at two coal-fired power plants in Russia. The first project has modified the wet particulate matter (PM scrubber installed in Toliatti thermal plant to allow for addition of chemical reagents (oxidants into the closedloop liquid spray system. The addition of oxidant resulted in significant improvement of mercury capture from 20% total mercury removal (without the additive up to 60% removal (with the additive. It demonstrates the effectiveness of sorbent injection technologies in conjunction with an electrostatic precipitator (ESP. ESPs are installed at 60%, while wet PM scrubbers are installed at 30% of total coal-fired capacity in Russia. Thus, the two UNEP Coal Partnership projects address the majority of PM emission control configurations occurring in Russia.

  12. Phenomenology and course of severe accidents in PWR-plants training by teaching and demonstration

    International Nuclear Information System (INIS)

    Sonnenkalb, M.; Rohde, J.

    1999-01-01

    A special one day training course on 'Phenomenology and Course of Severe Accidents in PWR-Plants' was developed at GRS initiated by the interest of German utilities. The work was done in the frame of projects sponsored by the German Ministries for Environment, Nature Conservation and Nuclear Safety (BMW) and for Education, Science, Research and Technology (BMBF). In the paper the intention and the subject of this training course are discussed and selected parts of the training course are presented. Demonstrations are made within this training course with the GRS simulator system ATLAS to achieve a broader understanding of the phenomena discussed and the propagation of severe accidents on a plant specific basis. The GRS simulator system ATLAS is linked in this case to the integral code MELCOR and pre-calculated plant specific severe accident calculations are used for the demonstration together with special graphics showing plant specific details. Several training courses have been held since the first one in November, 1996 especially to operators, shift personal and the management board of a German PWR. In the meantime the training course was updated and suggestions for improvements from the participants were included. In the future this training course will be made available for members of crisis teams, instructors of commercial training centres and researchers of different institutions too. (author)

  13. Design demonstrations for the remaining 19 Category B tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1993-06-01

    This document presents design demonstrations conducted of liquid low-level waste (LLLW) storage tank systems located at the Oak Ridge National Laboratory (ORNL). ORNL has conducted research in energy related fields since 1943. The facilities used to conduct the research include nuclear reactors, chemical pilot plants, research laboratories, radioisotope production laboratories, and support facilities. These facilities have produced a variety of radioactive and/or hazardous wastes that have been transported and stored through an extensive network of piping and tankage. Demonstration of the design of these tank systems has been stipulated by the Federal Facility Agreement (FFA) between the EPA (United States Environmental Protection Agency)-Region IV; the Tennessee Department of Environment and Conservation (TDEC); and the DOE. The FFA establishes four categories of tank systems: Category A-New or Replacement Tank Systems with Secondary Containment; Category B-Existing Tank Systems with Secondary Containment; Category C-Existing Tank Systems Without Secondary Containment, and Category D-Existing Tank Systems Without Secondary Containment That are Removed from Service. This document provides a design demonstration of the secondary containment and ancillary equipment of 19 tank systems listed in the FFA as Category B. The design demonstration for each tank is presented in Section 2. The assessments assume that each tank system was constructed in accordance with the design drawings and construction specifications for that system unless specified otherwise. Each design demonstration addresses system conformance to the requirements of the FFA (Appendix F, Section C)

  14. Recuperative aluminium recycling plant. A demonstration at J. McIntyre (Aluminium) Ltd. [Nottingham (GB)

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    Direct energy savings worth up to 470,000 pounds/year are being achieved by J McIntyre (Aluminium) Ltd in the United Kingdom as a result of the development of a recuperative aluminium recycling plant. The overall design incorporates a novel version of a closed-well furnace coupled with a radically improved design of dry hearth furnace. The plant not only treats clean scrap more efficiently than at present, but will also treat contaminated scrap which has not previously been recycled in an environmentally acceptable way. This is because the plant incorporates fume pyrolysis, afterburning of organics, recuperation and fume treatment. At 1987 prices the total installed plant cost was 1.3M.pounds. The direct energy saving at 1987 fuel prices was between 294,000 and 470,000pounds/year. Also, the improved melting technique has reduced metal lost as dross by 2 - 8% (420 -1,680 tonnes/year) when compared to other furnace operations. The improved metal recovery (1987 prices) was worth a further 400,000 pounds - 1,600,000 pounds. Taking median figures for the total fuel-plus-metal savings results in a payback on the project of only 14 months, some six months less than anticipated. Other consequential benefits which have also helped in reducing operating costs have been improvements in output per man, reductions in sickness and absenteeism, and reduced down-time for maintenance and repair. (author).

  15. Experience of upgrading existing Russian designed nuclear plants

    International Nuclear Information System (INIS)

    Yanev, P.I.; Facer, R.I.

    1993-01-01

    From the reviewed experiences of upgrading existing Russian designed nuclear plants both of WWER and RBMK type the conclusions drawn are as follows. For the countries operating Russian designed plants it is necessary to adopt a pragmatic approach where all changes must be demonstrated to improve the safety of the plant and safety must be demonstrably improving. Care must be taken to avoid the pitfalls of excessive regulatory demands which are not satisfied and the development of an attitude of disregarding requirements on the basis that they are not enforced. For the lending countries and organizations, it is necessary to ensure that assistance is given to the operating organizations so that the most effective use of funds can be achieved. The experience in the West is that over-regulation and excessive expenditure do not necessarily lead to improved safety. They can lead to significant waste of resources. The use of western technology is recommended but where it is necessary and where it provides the greatest benefit

  16. Design of a lunar oxygen production plant

    Science.gov (United States)

    Radhakrishnan, Ramalingam

    1990-01-01

    To achieve permanent human presence and activity on the moon, oxygen is required for both life support and propulsion. Lunar oxygen production using resources existing on the moon will reduce or eliminate the need to transport liquid oxygen from earth. In addition, the co-products of oxygen production will provide metals, structural ceramics, and other volatile compounds. This will enable development of even greater self-sufficiency as the lunar outpost evolves. Ilmenite is the most abundant metal-oxide mineral in the lunar regolith. A process involving the reaction of ilmenite with hydrogen at 1000 C to produce water, followed by the electrolysis of this water to provide oxygen and recycle the hydrogen has been explored. The objective of this 1990 Summer Faculty Project was to design a lunar oxygen-production plant to provide 5 metric tons of liquid oxygen per year from lunar soil. The results of this study describe the size and mass of the equipment, the power needs, feedstock quantity and the engineering details of the plant.

  17. Integral design small nuclear power plant UNITHERM

    International Nuclear Information System (INIS)

    Adamovich, L. A.; Grechko, G. I.; Ulasevich, V. K.; Shishkin, V. A.

    1995-01-01

    The need to erect expensive energy transmission lines to these places demands to use independent local energy sources. Therefore, a reasonable alternative to the plants fired fossil fuel, mostly hydrocarbon fuel, may come from the nuclear power plants (NPP) of relatively small capacity which are nonattended, shipped to the site by large-assembled modules and completely withdrawable from the site during decommissioning. Application of NPPs for power and heat supply may prove to be cost-efficient and rather positive from social and ecological point of view. UNITHERM NPP belongs to such energy sources and may be used for heat and power supply. Heat can be provided both as hot water and superheated steam. The consumers are able to specify heat/energy supply ratio. NPP design provides for independent energy supply to the consumers and the possibility to disconnect each of them without disruption of operation of the others. Thermal hydraulic diagram of UNITHERM NPP provides for the use of three interconnected, process circuits. The consumers of thermal energy (turbogenerator unit and boilers of the central heating unit) are arranged in the last circuit

  18. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    International Nuclear Information System (INIS)

    Tippets, F.E.

    1975-01-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  19. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Tippets, F E

    1975-07-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  20. Quality assurance in the design of nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    This Safety Guide provides the requirements and recommendations related to the establishment and implementation of quality assurance for design of items for a nuclear power plant. The requirements of this Guide shall be applied to the extent necessary during all constituent activities of the nuclear power plant project, such as design, manufacture, construction, commissioning and operations. Its requirements and recommendations shall be implemented, as appropriate, by the responsible organization or by its designated representatives: by plant designers, architect-engineers or manufacturers, when involved in performing design activities related to items to be manufactured; by site constructors, when involved in field engineering activities; by plant operators and other organizations, when involved in design activities related to plant modifications or to selection of spare or replacement parts; and by design consultants and other technical organizations, when performing any engineering activity that affects the work of other design organizations during various stages of nuclear power plant projects

  1. Design demonstrations for Category B tank system piping at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1994-02-01

    Demonstration of the design of the tank systems described in this report is stipulated by the Federal Facility Agreement (FFA) between the U.S. Environmental Protection Agency-Region IV, the Tennessee Department of Environment and Conservation, and the U.S. Department of Energy. This report provides a design demonstration of the secondary containment and ancillary equipment of 30 piping systems listed in the FFA as Category B (i.e., existing tank systems with secondary containment). The design demonstrations were developed using information obtained from design drawings (as-built when available), construction specifications, and interviews with facility operators. Each design demonstration addresses system conformance to the requirements of the FFA (Appendix F, Section C). Deficiencies or restrictions regarding the ability to demonstrate that each of the containment systems conforms to FFA requirements are noted in the discussion of each piping system

  2. The design of operating procedures manuals for nuclear power plants

    International Nuclear Information System (INIS)

    Bohr, E.; Preuss, W.; Reinartz, G.; Thau, G.

    1977-03-01

    This report describes the findings of a research on the desirable design of operating procedures manuals for nuclear power plants. The work was supported by a grant of the Federal Department of the Interior. Information was acquired from different sources. Interviews and discussions on manual design were carried out with manual users in nuclear power plants. Moreover, tasks carried out using procedures were either observed or, alternatively, the manner of using procedures was elicited by interviews. In addition, manual writers, managers from manufacturers and utilities, nuclear experts, and individuals involved in manual specification activities were interviewed. A major source of information has been the pertinent scientific and technical findings scattered in the literature on topics such as instructional technology, engineering psychology, psycholinguistics, and typography. A comprehensive bibliography is included. General rules are established on designing instructional material for use on the job, aiming at increasing their legability, comprehensibility, and suitability to guide human performance. The application of these rules to the design of individual operating procedures is demonstrated. Recommendations are given on the design, layout, development and implementation of manuals. (orig.) [de

  3. Field demonstration of ex situ biological treatability of contaminated groundwater at the Strachan gas plant

    International Nuclear Information System (INIS)

    Kurz, M.D.; Stepan, D.J.

    1997-03-01

    A multi-phase study was conducted to deal with the issues of groundwater and soil contamination by sour gas processing plants in Alberta. Phase One consisted of a review of all soil and groundwater monitoring data submitted to Alberta Environment by sour gas plants in accordance with the Canadian Clean Water Act. The current phase involves the development, evaluation and demonstration of selected remediation technologies to address subsurface contamination of sediments and groundwater at sour gas treatment plants with special attention to the presence of natural gas condensate in the subsurface. Results are presented from a pilot-scale biological treatability test that was performed at the Gulf Strachan Natural Gas Processing Plant in Rocky Mountain House, Alberta, where contaminated groundwater from the plant was being pumped to the surface through many recovery wells to control contaminant migration. The recovered groundwater was directed to a pump-and-treat system that consisted of oil-water separation, iron removal, hardness removal, and air stripping, before being reinjected. The pilot-scale biological treatability testing was conducted to evaluate process stability in treating groundwater without pretreatment for iron and hardness reduction and to evaluate the removal of organic contaminants. Results of a groundwater characterization analysis are discussed. Chemical characteristics of the groundwater at the Strachan Gas Plant showed that an ex situ remediation technology would address the dissolved volatile and semi-volatile organic contamination from natural gas condensates, as well as the nitrogenous compounds resulting from the use of amine-based process chemicals. 4 refs., 5 tabs., 4 figs

  4. Advanced Neutron Source: Plant Design Requirements

    International Nuclear Information System (INIS)

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS

  5. Round Robin Test for Performance Demonstration System of Ultrasound Examination Personnel in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Young Ho; Yang, Seung Han; Kim, Yong Sik; Yoon, Byung Sik; Lee, Hee Jong

    2005-01-01

    Ultrasound testing performance during in-service inspection for the main components of NPPs is strongly affected by each examination person. Therefore, ASME established a more strict qualification requirement in Sec. XI Appendix VIII for the ultrasound testing personnel in nuclear power plants. The Korean Performance Demonstration (KPD) System according to the ASME code for the ultrasonic testing personnel, equipments, and procedures to apply to the Class 1 and 2 piping ultrasound examination of nuclear power plants in Korea was established. And a round robin test was conducted in order to verify the effectiveness of PD method by comparing the examination results from the method of Performance Demonstration (PD) and a traditional ASME code dB-drop method. The round robin test shows that the reliability of the PD method is better than that of the dB-drop method. As a result, application of the PD method to the in-service inspection of the nuclear power plants will improve the performance of ultrasound testing

  6. Demonstration of the economic feasibility of plant tissue culture for jojoba (Simmondsia chinensis) and Euphorbia spp

    Energy Technology Data Exchange (ETDEWEB)

    Sluis, C.

    1980-09-01

    The economic feasibility of plant tissue culture was demonstrated as applied to two plants: jojoba (Simmondsia chinensis) and Euphorbia spp. The gopher weed (Euphorbia lathyris) was selected as the species of Euphorbia to research due to the interest in this plant as a potential source of hydrocarbon-like compounds. High yield female selections of jojoba were chosen from native stands and were researched to determine the economic feasibility of mass producing these plants via a tissue culture micropropagation program. The female jojoba selection was successfully mass produced through tissue culture. Modifications in initiation techniques, as well as in multiplication media and rooting parameters, were necessary to apply the tissue culture system, which had been developed for juvenile seedling tissue, to mature jojobas. Since prior attempts at transfer of tissue cultured plantlets were unsuccessful, transfer research was a major part of the project and has resulted in a system for transfer of rooted jojoba plantlets to soil. Euphorbia lathyris was successfully cultured using shoot tip cultures. Media and procedures were established for culture initiation, multiplication of shoots, callus induction and growth, and root initiation. Well-developed root systems were not attained and root initiation percentages should be increased if the system is to become commercially feasible.

  7. How to optimize hydrogen plant designs

    Energy Technology Data Exchange (ETDEWEB)

    van Weenen, W F; Tielrooy, J

    1983-01-01

    In a typical hydrogen plant of the type which will be discussed, methane or higher hydrocarbons are reformed with steam in a steam hydrocarbon reformer operating at a pressure of 250 to 400 psig, a temperature of 1500 to 1600/sup 0/F, and with a ratio of steam to carbon in the feed of about 3.0. Following the reformer and cooling, there is a single stage of high temperature carbon monoxide shift conversion. Optionally, after further cooling, this may be followed by a second stage of carbon monoxide shift conversion operating at a lower temperature to obtain a more favourable equilibrium; this is called low temperature shift conversion. After cooling to ambient temperature, and separation of the condensate, the gas is passed through a Pressure Swing Adsorption (PSA)l unit which removes all the impurities along with a small amount of hydrogen. The waste gas from the PSA unit containing all the impurities is used as fuel to the reformer. Heat is recovered from the reformer flue gas, reformer product, high temperature shift converter product and low temperature shift converter product. This paper discusses some of the process variables and design variables which must be considered in arriving at an optimized design. Seven different flow schemes are discussed in the light of the objectives they are designed for. The seven schemes and their objectives are: Flow Scheme 1 - lowest first cost; moderate efficiency, Flow Scheme 2 - high efficiency, low cost; Flow Scheme 3 - low feed plus fuel, moderately high efficiency; Flow Scheme 4 - lowest feed plus fuel; Flow Scheme 5 - lowest feed, low fuel; Flow Scheme 6 -lowest feed, highest efficiency; and Flow Scheme 7 - lowest feed plus fuel, export electric power instead of export electric power instead of export steam. 15 figures, 1 table.

  8. The National Long Term Care Demonstration: operational issues encountered in developing the research design.

    Science.gov (United States)

    Carcagno, G J; Kemper, P

    1983-01-01

    This paper describes the design of the National Long Term Care Demonstration and its evaluation and discusses a number of operational issues encountered in the design process: simultaneous design of research and operations, identification of the target population, randomization, collection of comparable data, development of an assessment instrument, potential changes in existing programs, and termination planning.

  9. Demonstrating Empathy: A Phenomenological Study of Instructional Designers Making Instructional Strategy Decisions for Adult Learners

    Science.gov (United States)

    Vann, Linda S.

    2017-01-01

    Instructional designers are tasked with making instructional strategy decisions to facilitate achievement of learning outcomes as part of their professional responsibilities. While the instructional design process includes learner analysis, that analysis alone does not embody opportunities to assist instructional designers with demonstrations of…

  10. Fuel Gas Demonstration Plant Program: Small-Scale Industrial Project. Environmental assessment statement

    Energy Technology Data Exchange (ETDEWEB)

    1979-02-01

    Solid, liquid, and gaseous by-products and wastes are generated during coal storage and processing, gasification, and gas cleanup. Recovery systems have been designed to collect and utilize by-products. Wastes will be placed in storage areas designed to prevent release of the materials to the environment. The coal gasification plant along with the solid waste disposal area will occupy approximately 115 acres. To prevent, to the fullest extent possible, degradation of groundwater and surface water resources, the coal stockpile, landfill, collection pond, settling basin, and drainage ditches will be constructed to prevent the seepage of potential contaminants into groundwater or the drainage of runoff into surface waters. Cooling water is the primary water requirement of the project. None of the water utilized in the gasification plant will be released into the area surface water system, but will be either recycled or directed into the settling basin. The gasification facility has the potential of emitting a broad spectrum of pollutants into the atmosphere. However, effective emission control procedures such as off-gas recycling, hydrogen sulfide removal, particulate removal, and flaring will be applied to minimize the plant's emissions. The necessity of monitoring the more exotic pollutants such as acid gases, trace elements, metal carbonyls, and a multitude of organic compounds, will be determined as the gasification facility becomes more of a reality and the latest literature and research developments can be surveyed to evaluate the emission rates, biological significance, and monitoring techniques for these pollutants.

  11. Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

    International Nuclear Information System (INIS)

    Lear, G.; Erickson, P.B.

    1975-01-01

    Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

  12. RIMAP demonstration project. Risk-based life management of piping system in power plant Heilbronn

    International Nuclear Information System (INIS)

    Bareiss, J.; Buck, P.; Matschecko, B.; Jovanovic, A.; Balos, D.; Perunicic, M.

    2004-01-01

    In the framework of EU project RIMAP [Risk Based Inspection and Maintenance Procedures for European Industry (2000)] a new European Guideline for optimized risk based maintenance and inspection planning of industrial plants (RBLM, Risk Based Life Management) is being developed. The RIMAP project consists of the three clustered projects: - development (RTD); - demonstration (DEMO): - thematic network (TN). Current work and future, planned work in RIMAP demonstration project on applications of the RIMAP methodology in power plants are presented briefly in the first part of the paper. Also presented in the paper are the results of a preliminary analysis of piping system in power plant Heilbronn using the concept of risk-based monitoring as part of overall concept of risk-based life management. Shortly the following issues are discussed in the paper: - identification of critical components; - application of a multilevel risk analysis (...from 'screening' to 'detailed analysis'); - determination of PoF (Probability of Failure); - determination of CoF (Consequence of Failure); - optimation of inspection and maintenance plan. From our experience with the application of the RIMAP methodology the following conclusions can be drawn: The use of risk-based methods in inspection and maintenance of piping systems in power plants gives transparency to the decision making process and gives an optimized maintenance policy based on current state of the components. The results of the work clearly show the power of the proposed method for concentration on critical items: out of 64 monitored components 5 were selected for intermediate analysis and only 1 for the detailed analysis (probabilistic high temperature fracture mechanics)

  13. The Blend Down Monitoring System Demonstration at the Padijcah Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Benton, J.; Close, D.; Johnson, W. Jr.; Kerr, P.; March-Leuba, J.; Mastal, E.; Moss, C.; Powell, D.; Sumner, J.; Uckan, T.; Vines, R.; Wright, P.D.

    1999-01-01

    Agreements between the governments of the US and the Russian Federation for the US purchase of low enriched uranium (LEU) derived from highly enriched uranium (HEU) from dismantled Russian nuclear weapons calls for the establishment of transparency measures to provide confidence that nuclear nonproliferation goals are being met. To meet these transparency goals, the agreements call for the installation of nonintrusive US instruments to monitor the down blending of HEU to LEU. The Blend Down Monitoring System (BDMS) has been jointly developed by the Los Alamos National Laboratory (LANL) and the Oak Ridge National Laboratory (ORNL) to continuously monitor 235 U enrichments and mass flow rates at Russian blending facilities. Prior to its installation in Russian facilities, the BDMS was installed and operated in a UF 6 flow loop in the Paducah Gaseous Diffusion Plant simulating flow and enrichment conditions expected in a typical down-blending facility. A Russian delegation to the US witnessed the equipment demonstration in June, 1998. To conduct the demonstration in the Paducah Gaseous Diffusion Plant (PGDP), the BDMS was required to meet stringent Nuclear Regulatory Commission licensing, safety and operational requirements. The Paducah demonstration was an important milestone in achieving the operational certification for the BDMS use in Russian facilities

  14. Osiris and SOMBRERO inertial confinement fusion power plant designs

    International Nuclear Information System (INIS)

    Meier, W.R.; Bieri, R.L.; Monsler, M.J.

    1992-03-01

    Conceptual designs and assessments have been completed for two inertial fusion energy (IFE) electric power plants. The detailed designs and results of the assessment studies are presented in this report. Osiris is a heavy-ion-beam (HIB) driven power plant and SOMBRERO is a Krypton-Fluoride (KrF) laser-driven power plant. Both plants are sized for a net electric power of 1000 MWe

  15. Design demonstrations for Category B tank systems piping at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1994-12-01

    Demonstration of the design of the piping systems described in this report is stipulated by the Federal Facility Agreement (FFA) between the U.S. Environmental Protection Agency (EPA)-Region IV, the Tennessee Department of Environment and Conservation (TDEC), and the U.S. Department of Energy (DOE). This report provides a design demonstration of the secondary containment and ancillary equipment of 30 piping systems designated in the FFA as Category B (i.e., existing tank systems with secondary containment). Based on the findings of the Design Demonstrations for the Remaining 19 Category B Tank Systems, (DOE/OR/03-1150 ampersand D2), three tank systems originally designated as Category B have been redesignated as Category C (i.e., existing tank systems without secondary containment). The design demonstrations were developed using information obtained from design drawings (as-built when available), construction specifications, and interviews with facility operators. Each design demonstration addresses system conformance to the requirements of the FFA (Appendix F, Section C). Deficiencies or restrictions regarding the ability to demonstrate that each of the containment systems conforms to FFA requirements are noted in the discussion of each piping system and presented in Table 2.0-1

  16. Nuclear power plant equipment design and construction rules

    International Nuclear Information System (INIS)

    Boiron, P.

    1983-03-01

    Presentation of the AFCEN (French association for nuclear power plant equipment design and construction rules) working, of its edition activity and of somes of its edited documents such as RCC-C (design and construction rules for PWR power plant fuel assemblies) and RCC-E (design and construction rules for nuclear facility electrical equipments) [fr

  17. SRC-I demonstration plant analytical laboratory methods manual. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Klusaritz, M.L.; Tewari, K.C.; Tiedge, W.F.; Skinner, R.W.; Znaimer, S.

    1983-03-01

    This manual is a compilation of analytical procedures required for operation of a Solvent-Refined Coal (SRC-I) demonstration or commercial plant. Each method reproduced in full includes a detailed procedure, a list of equipment and reagents, safety precautions, and, where possible, a precision statement. Procedures for the laboratory's environmental and industrial hygiene modules are not included. Required American Society for Testing and Materials (ASTM) methods are cited, and ICRC's suggested modifications to these methods for handling coal-derived products are provided.

  18. Application and demonstration of oxyfuel combustion technologies to the existing power plant in Australia

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Terutoshi; Yamada, Toshihiko; Watanabe, Shuzo; Kiga, Takashi; Gotou, Takahiro [IHI Corporation, Tokyo (Japan). Power Plant Div.; Misawa, Nobuhiro [Electric Power Development Co., Ltd., Tokyo (Japan); Spero, Chris [CS Energy Ltd, Brisbane (Australia)

    2013-07-01

    Oxyfuel combustion is able to directly make the highly concentrated CO{sub 2} from the flue gas of pulverized coal fired power plant and, therefore, is expected as one of the promising technologies for CO{sub 2} capture. We are advancing the Oxyfuel combustion demonstration project, which is called Callide Oxyfuel Project, with the support of both Australian and Japanese governments. Currently the boiler retrofit work is completed and the commissioning in air combustion is going on. In this paper, we introduce the general outline of the Callide Oxyfuel Project and its progress.

  19. Instrumentation and process control for fossil demonstration plants. Quarterly technical progress report, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    LeSage, L.G.

    1977-07-01

    Work has been performed on updating the study of the state-of-the-art of instrumentation for Fossil Demonstration Plants (FDP), development of mass-flow and other on-line instruments for FDP, process control analysis for FDP, and organization of a symposium on instrumentation and control for FDP. A Solids/Gas Flow Test Facility (S/GFTF) under construction for instrument development, testing, evaluation, and calibration is described. The development work for several mass-flow and other on-line instruments is described: acoustic flowmeter, capacitive density flowmeter, neutron activation flowmeter and composition analysis system, gamma ray correlation flowmeter, optical flowmeter, and capacitive liquid interface level meter.

  20. Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

    International Nuclear Information System (INIS)

    Swannack, D.L.

    1982-01-01

    This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

  1. Design and construction of an inexpensive homemade plant growth chamber.

    Science.gov (United States)

    Katagiri, Fumiaki; Canelon-Suarez, Dario; Griffin, Kelsey; Petersen, John; Meyer, Rachel K; Siegle, Megan; Mase, Keisuke

    2015-01-01

    Plant growth chambers produce controlled environments, which are crucial in making reproducible observations in experimental plant biology research. Commercial plant growth chambers can provide precise controls of environmental parameters, such as temperature, humidity, and light cycle, and the capability via complex programming to regulate these environmental parameters. But they are expensive. The high cost of maintaining a controlled growth environment is often a limiting factor when determining experiment size and feasibility. To overcome the limitation of commercial growth chambers, we designed and constructed an inexpensive plant growth chamber with consumer products for a material cost of $2,300. For a comparable growth space, a commercial plant growth chamber could cost $40,000 or more. Our plant growth chamber had outside dimensions of 1.5 m (W) x 1.8 m (D) x 2 m (H), providing a total growth area of 4.5 m2 with 40-cm high clearance. The dimensions of the growth area and height can be flexibly changed. Fluorescent lights with large reflectors provided a relatively spatially uniform photosynthetically active radiation intensity of 140-250 μmoles/m2/sec. A portable air conditioner provided an ample cooling capacity, and a cooling water mister acted as a powerful humidifier. Temperature, relative humidity, and light cycle inside the chamber were controlled via a z-wave home automation system, which allowed the environmental parameters to be monitored and programmed through the internet. In our setting, the temperature was tightly controlled: 22.2°C±0.8°C. The one-hour average relative humidity was maintained at 75%±7% with short spikes up to ±15%. Using the interaction between Arabidopsis and one of its bacterial pathogens as a test experimental system, we demonstrate that experimental results produced in our chamber were highly comparable to those obtained in a commercial growth chamber. In summary, our design of an inexpensive plant growth chamber

  2. Design and construction of an inexpensive homemade plant growth chamber.

    Directory of Open Access Journals (Sweden)

    Fumiaki Katagiri

    Full Text Available Plant growth chambers produce controlled environments, which are crucial in making reproducible observations in experimental plant biology research. Commercial plant growth chambers can provide precise controls of environmental parameters, such as temperature, humidity, and light cycle, and the capability via complex programming to regulate these environmental parameters. But they are expensive. The high cost of maintaining a controlled growth environment is often a limiting factor when determining experiment size and feasibility. To overcome the limitation of commercial growth chambers, we designed and constructed an inexpensive plant growth chamber with consumer products for a material cost of $2,300. For a comparable growth space, a commercial plant growth chamber could cost $40,000 or more. Our plant growth chamber had outside dimensions of 1.5 m (W x 1.8 m (D x 2 m (H, providing a total growth area of 4.5 m2 with 40-cm high clearance. The dimensions of the growth area and height can be flexibly changed. Fluorescent lights with large reflectors provided a relatively spatially uniform photosynthetically active radiation intensity of 140-250 μmoles/m2/sec. A portable air conditioner provided an ample cooling capacity, and a cooling water mister acted as a powerful humidifier. Temperature, relative humidity, and light cycle inside the chamber were controlled via a z-wave home automation system, which allowed the environmental parameters to be monitored and programmed through the internet. In our setting, the temperature was tightly controlled: 22.2°C±0.8°C. The one-hour average relative humidity was maintained at 75%±7% with short spikes up to ±15%. Using the interaction between Arabidopsis and one of its bacterial pathogens as a test experimental system, we demonstrate that experimental results produced in our chamber were highly comparable to those obtained in a commercial growth chamber. In summary, our design of an inexpensive plant

  3. Design and construction of nuclear power plants

    International Nuclear Information System (INIS)

    Meiswinkel, Ruediger; Meyer, Julian; Schnell, Juergen

    2013-01-01

    Despite all the efforts being put into expanding renewable energy sources, large-scale power stations will be essential as part of a reliable energy supply strategy for a longer period. Given that they are low on CO2 emissions, many countries are moving into or expanding nuclear energy to cover their baseload supply. Building structures required for nuclear installations whose protective function means they are classified as safety-related, have to meet particular construction requirements more stringent than those involved in conventional construction. This book gives a comprehensive overview from approval aspects given by nuclear and construction law, with special attention to the interface between plant and construction engineering, to a building structure classification. All life cycle phases are considered, with the primary focus on execution. Accidental actions on structures, the safety concept and design and fastening systems are exposed to a particular treatment. Selected chapters of the German concrete yearbook ''Beton-Kalender'' are now available in English. The new English BetonKalender Series delivers internationally useful engineering expertise and industrial know-how from Germany.

  4. Cardinal principle and application practice of 3D digital model design for nuclear power plant

    International Nuclear Information System (INIS)

    Wang Ruobing; Wu Yan

    2005-01-01

    The practical application of 3D digital model design at nuclear power plants was introduced in detail in the paper. The whole process for system choice, program constitution, model design and project practice were also summarized. By demonstrating the cardinal principal and application practice of 3D digital model design as an important sub-project of CGNPC Digital Plant, the paper validates the rationality and validity of the major architecture system and program configuration of the digital plant, carries out beneficial attempt and study in the overall power plant life engineering management and site practice, and has achieved significant engineering and social benefits. The success of practices in the project accelerates the extended and extensive application of Digital Plant in the operation and maintenance simulation of Daya Bay and Ling'ao Nuclear Power Plants, and the engineering design management for Ling'ao II and III of CGNPC on a consolidated basis. (authors)

  5. System 80+ integrated design of a complete plant

    International Nuclear Information System (INIS)

    Turk, R.S.; Stamm, S.L.; Fox, W.A.

    1992-01-01

    In 1985, ABB-Combustion Engineering Nuclear Power (ABB-CENP) and elements of Duke Power Company [now Duke Engineering ampersand Services (DE ampersand S)] joined forces under the aegis of the Electric Power Research Institute (EPRI) Advanced Light Water Reactor (ALWR) Program to develop, with the sponsoring utilities, the design requirements for the next generation of nuclear power plants. With support from the US Department of Energy, ABB-CENP and DE ampersand S again teamed up the following year to initiate a project to design and license the System 80+ standard plant design, an advanced pressurized water reactor that meets these utility requirements. A distinguishing feature of the System 80+ standard design is that it is an essentially complete plant, predesigned and prelicensed to ensure rapid and economical construction. This is in stark contrast to typical prior conduct, where the reactor vendor offered only the nuclear steam supply system and the plant was built on a design-as-you-go basis with constant pressure to release individual elements of the plant design for construction or procurement as soon as possible. Now, however, the design process can be integrated over the total plant, ensuring that the goals set for ALWRs can be met. This integrated design process is manifested in several ways: (1) broad-based participation during the design process by involving designers, analysts, suppliers, constructors, and operators; (2) use of probabilistic risk assessment (PRA) as a design tool to aid in evaluating design features on a total-plant basis; (3) application of human factors engineering methods to a total plant distributed control system to improve the human-machine interface in the design; and (4) use of computer-aided design to enhance assessment of interactions and impacts of all aspects of the total plant. Each of these aspects of integrated plant design is discussed in this paper

  6. Design demonstrations for category B tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    1994-11-01

    This document presents design demonstrations conducted of liquid low-level waste (LLLW) storage tank systems located at the Oak Ridge National Laboratory (ORNL). Demonstration of the design of these tank systems has been stipulated by the Federal Facility Agreement (FFA) between the US Environmental Protection Agency (EPA)-Region IV; the Tennessee Department of Environment and Conservation (TDEC); and the DOE. The FFA establishes four categories of tanks. These are: Category A -- New or replacement tank systems with secondary containment; Category B -- Existing tank systems with secondary containment; Category C -- Existing tank systems without secondary containment; Category D -- Existing tank systems without secondary containment that are removed from service. This document provides a design demonstration of the secondary containment and ancillary equipment of 11 tank systems listed in the FFA as Category B. The design demonstration for each tank is presented.

  7. Design demonstrations for category B tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1994-11-01

    This document presents design demonstrations conducted of liquid low-level waste (LLLW) storage tank systems located at the Oak Ridge National Laboratory (ORNL). Demonstration of the design of these tank systems has been stipulated by the Federal Facility Agreement (FFA) between the US Environmental Protection Agency (EPA)-Region IV; the Tennessee Department of Environment and Conservation (TDEC); and the DOE. The FFA establishes four categories of tanks. These are: Category A -- New or replacement tank systems with secondary containment; Category B -- Existing tank systems with secondary containment; Category C -- Existing tank systems without secondary containment; Category D -- Existing tank systems without secondary containment that are removed from service. This document provides a design demonstration of the secondary containment and ancillary equipment of 11 tank systems listed in the FFA as Category B. The design demonstration for each tank is presented

  8. The research for the design verification of nuclear power plant based on VR dynamic plant

    International Nuclear Information System (INIS)

    Wang Yong; Yu Xiao

    2015-01-01

    This paper studies a new method of design verification through the VR plant, in order to perform verification and validation the design of plant conform to the requirements of accident emergency. The VR dynamic plant is established by 3D design model and digital maps that composed of GIS system and indoor maps, and driven by the analyze data of design analyzer. The VR plant could present the operation conditions and accident conditions of power plant. This paper simulates the execution of accident procedures, the development of accidents, the evacuation planning of people and so on, based on VR dynamic plant, and ensure that the plant design will not cause bad effect. Besides design verification, simulated result also can be used for optimization of the accident emergency plan, the training of accident plan and emergency accident treatment. (author)

  9. Prototype CIRCE plant-industrial demonstration of heavy-water production from a reformed hydrogen source

    Energy Technology Data Exchange (ETDEWEB)

    Spagnolo, D.A.; Boniface, H.A.; Sadhankar, R.R.; Everatt, A.E.; Miller, A.I. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Blouin, J. [Air Liquide Canada, Hamilton, Ontario (Canada)

    2002-09-01

    Heavy-water (D{sub 2}0) production has been dominated by the Girdler-Sulphide (G-S) process, which suffers several intrinsic disadvantages that lead to high production costs. Processes based on hydrogen/water exchange have become more attractive with the development of proprietary wetproofed catalysts by AECL. One process that is synergistic with industrial hydrogen production by steam methane reforming (SMR), the combined industrial reforming and catalytic exchange (CIRCE) process, offers the best prospect for commercialization. SMRs are common globally in the oil upgrading and ammonia industries. To study the CIRCE process in detail, AECL, in collaboration with Air Liquide Canada, constructed a prototype CIRCE plant (PCP) in Hamilton, ON. The plant became fully operational in 2000 July and is expected to operate to at least the late fall of 2002. To date, plant operation has confirmed the adequacy of the design and the capability of enriching deuterium to produce heavy water without compromising hydrogen production. The proprietary wetproofed catalyst has performed as expected, both in activity and in robustness. (author)

  10. Prototype CIRCE plant-industrial demonstration of heavy-water production from a reformed hydrogen source

    International Nuclear Information System (INIS)

    Spagnolo, D.A.; Boniface, H.A.; Sadhankar, R.R.; Everatt, A.E.; Miller, A.I.; Blouin, J.

    2002-09-01

    Heavy-water (D 2 0) production has been dominated by the Girdler-Sulphide (G-S) process, which suffers several intrinsic disadvantages that lead to high production costs. Processes based on hydrogen/water exchange have become more attractive with the development of proprietary wetproofed catalysts by AECL. One process that is synergistic with industrial hydrogen production by steam methane reforming (SMR), the combined industrial reforming and catalytic exchange (CIRCE) process, offers the best prospect for commercialization. SMRs are common globally in the oil upgrading and ammonia industries. To study the CIRCE process in detail, AECL, in collaboration with Air Liquide Canada, constructed a prototype CIRCE plant (PCP) in Hamilton, ON. The plant became fully operational in 2000 July and is expected to operate to at least the late fall of 2002. To date, plant operation has confirmed the adequacy of the design and the capability of enriching deuterium to produce heavy water without compromising hydrogen production. The proprietary wetproofed catalyst has performed as expected, both in activity and in robustness. (author)

  11. Prototype CIRCE plant - industrial demonstration of heavy water production from reformed hydrogen source

    International Nuclear Information System (INIS)

    Spagnolo, D.A.; Boniface, H.A.; Sadhankar, R.R.; Everatt, A.E.; Miller, A.I.; Blouin, J.

    2002-01-01

    Heavy water (D 2 0) production has been dominated by the Girdler-Sulphide (G-S) process, which suffers several intrinsic disadvantages that lead to high production costs. Processes based on hydrogen/water exchange have become more attractive with the development of proprietary wetproofed catalysts by AECL. One process that is synergistic with industrial hydrogen production by steam methane reforming (SMR), the Combined Industrial Reforming and Catalytic Exchange (CIRCE) process, offers the best prospect for commercialization. SMRs are common globally in the oil-upgrading and ammonia industries. To study the CIRCE process in detail, AECL, in collaboration with Air Liquide Canada, constructed a prototype CIRCE plant (PCP) in Hamilton, Ontario. The plant became fully operational in 2000 July and is expected to operate to at least late fall of 2002. To-date, plant operation has confirmed the adequacy of the design and the capability of enriching deuterium to produce heavy water without compromising hydrogen production. The proprietary wetproofed catalyst has performed as expected, both in activity and in robustness. (author)

  12. Preliminary Studies of the Demonstration Plant of the Energy Amplifier (EA) by LAESA

    International Nuclear Information System (INIS)

    Abanades, A.; Nunez-Lagos, R.; Perez-Navarro, A.; Rubio, J.A.

    1998-01-01

    On July 1997, an international company named LAESA has finished its foundational period, with the headquarters in Zaragoza. The main initial goal of the company is, based on the conceptual design of an Energy Amplifier from C. Rubbia and collaborators, to take the necessary steps for developing the engineering design and the construction of a demonstration unit of EA (approx 100 MWt in its first phase and 250 MWt in the second one). The company will be mainly charged of the technical aspects required by the construction of the prototype and its funding comes both from private and public capital, as well as the selling of services. The prototype intends to use a final accelerator current of approx 12 mA and a beam energy close to 400 MeV. The construction period is expected to be 5 to t years and the initial emphasis of the installation is to demonstrate its capability for a massive incineration of long lived radioactive waste, with the corresponding energy generation. The necessary development to prepare de engineering design is being the subject of a R and D programme in Spain. Other countries with or without companies inside LAESA are also carrying on other Rand D programmes. The expected planning foresees an integrated R and D programme, a first version of the engineering design and the initial contacts both with the Spanish Nuclear Regulatory Commission, as well as the Spanish Environmental authorities. During the Conference, the status of the engineering design will be shown. (Author) 3 refs

  13. Publishing Single-Case Research Design Studies That Do Not Demonstrate Experimental Control

    Science.gov (United States)

    Tincani, Matt; Travers, Jason

    2018-01-01

    Demonstration of experimental control is considered a hallmark of high-quality single-case research design (SCRD). Studies that fail to demonstrate experimental control may not be published because researchers are unwilling to submit these papers for publication and journals are unlikely to publish negative results (i.e., the file drawer effect).…

  14. Demonstration: SpaceExplorer - A Tool for Designing Ubiquitous Web Applications for Collections of Displays

    DEFF Research Database (Denmark)

    Hansen, Thomas Riisgaard

    2007-01-01

    This demonstration presents a simple browser plug-in that grant web applications the ability to use multiple nearby devices for displaying web content. A web page can e.g. be designed to present additional information on nearby devices. The demonstration introduces a light weight peer-to-peer arc...

  15. A Function-Behavior-State Approach to Designing Human Machine Interface for Nuclear Power Plant Operators

    Science.gov (United States)

    Lin, Y.; Zhang, W. J.

    2005-02-01

    This paper presents an approach to human-machine interface design for control room operators of nuclear power plants. The first step in designing an interface for a particular application is to determine information content that needs to be displayed. The design methodology for this step is called the interface design framework (called framework ). Several frameworks have been proposed for applications at varying levels, including process plants. However, none is based on the design and manufacture of a plant system for which the interface is designed. This paper presents an interface design framework which originates from design theory and methodology for general technical systems. Specifically, the framework is based on a set of core concepts of a function-behavior-state model originally proposed by the artificial intelligence research community and widely applied in the design research community. Benefits of this new framework include the provision of a model-based fault diagnosis facility, and the seamless integration of the design (manufacture, maintenance) of plants and the design of human-machine interfaces. The missing linkage between design and operation of a plant was one of the causes of the Three Mile Island nuclear reactor incident. A simulated plant system is presented to explain how to apply this framework in designing an interface. The resulting human-machine interface is discussed; specifically, several fault diagnosis examples are elaborated to demonstrate how this interface could support operators' fault diagnosis in an unanticipated situation.

  16. Nuclear power. Volume 1. Nuclear power plant design

    International Nuclear Information System (INIS)

    Pedersen, E.S.

    1978-01-01

    NUCLEAR POWER PLANT DESIGN is intended to be used as a working reference book for management, engineers and designers, and as a graduate-level text for engineering students. The book is designed to combine theory with practical nuclear power engineering and design experience, and to give the reader an up-to-date view of the status of nuclear power and a basic understanding of how nuclear power plants function. Volume 1 contains the following chapters; (1) nuclear reactor theory; (2) nuclear reactor design; (3) types of nuclear power plants; (4) licensing requirements; (5) shielding and personnel exposure; (6) containment and structural design; (7) main steam and turbine cycles; (8) plant electrical system; (9) plant instrumentation and control systems; (10) radioactive waste disposal (waste management) and (11) conclusion

  17. Safety principles and design management of Chashma Nuclear Power Plant

    International Nuclear Information System (INIS)

    Geng Qirui; Cheng Pingdong

    1997-01-01

    The basic safety consideration and detailed design principles in the design of Chashma Nuclear Power Plant is elaborated. The management within the frame setting up by 'safety culture' and 'quality culture'

  18. Design basis reconstitution and configuration management of nuclear power plants

    International Nuclear Information System (INIS)

    Smith, P.R.; Cook, R.A.

    1990-01-01

    This paper presents a method by which all plants (those under construction, as well as those in operation) can benefit from a formal, step-by-step approach to data capture, storage, and retrieval for use throughout the plant life cycle. To demonstrate the viability and in-practice success of a single three-dimensional Configuration Management database, the specific capabilities of one such system are discussed in detail: the Plant Configuration Management System (PLANT/CMS). (orig.)

  19. Baseload Nitrate Salt Central Receiver Power Plant Design Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Tilley, Drake [Abengoa Solar LLC, Lakewood, CO (United States); Kelly, Bruce [Abengoa Solar LLC, Lakewood, CO (United States); Burkholder, Frank [Abengoa Solar LLC, Lakewood, CO (United States)

    2014-12-12

    The objectives of the work were to demonstrate that a 100 MWe central receiver plant, using nitrate salt as the receiver coolant, thermal storage medium, and heat transport fluid in the steam generator, can 1) operate, at full load, for 6,400 hours each year using only solar energy, and 2) satisfy the DOE levelized energy cost goal of $0.09/kWhe (real 2009 $). To achieve these objectives the work incorporated a large range of tasks relating to many different aspects of a molten salt tower plant. The first Phase of the project focused on developing a baseline design for a Molten Salt Tower and validating areas for improvement. Tasks included a market study, receiver design, heat exchanger design, preliminary heliostat design, solar field optimization, baseline system design including PFDs and P&IDs and detailed cost estimate. The baseline plant met the initial goal of less than $0.14/kWhe, and reinforced the need to reduce costs in several key areas to reach the overall $0.09/kWhe goal. The major improvements identified from Phase I were: 1) higher temperature salt to improve cycle efficiency and reduce storage requirements, 2) an improved receiver coating to increase the efficiency of the receiver, 3) a large receiver design to maximize storage and meet the baseload hours objective, and 4) lower cost heliostat field. The second Phase of the project looked at advancing the baseline tower with the identified improvements and included key prototypes. To validate increasing the standard solar salt temperature to 600 °C a dynamic test was conducted at Sandia. The results ultimately proved the hypothesis incorrect and showed high oxide production and corrosion rates. The results lead to further testing of systems to mitigate the oxide production to be able to increase the salt temperature for a commercial plant. Foster Wheeler worked on the receiver design in both Phase I and Phase II looking at both design and lowering costs utilizing commercial fossil boiler

  20. 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.

  1. Design, Construction, Demonstration and Delivery of an Automated Narrow Gap Welding System.

    Science.gov (United States)

    1982-06-29

    DESIGN, CONSTRUCTION, DEMONSTRATION AND DELIVERY OF WE DA4I &NARROW GAP CONTRACT NO. NOOGOO-81-C-E923 TO DAVID TAYLOR NAVAL RESEARCH AND DEVELOPMENT...the automated * Narrow Gap welding process, is the narrow (3/8 - inch), square-butt joint *design. This narrow joint greatly reduces the volume of weld...AD-i45 495 DESIGN CONSTRUCTION DEMONSTRATION AiND DELIVERY OF RN 1/j AUrOMATED NARROW GAP WELDING SYSTEMI() CRC AUTOMATIC WELDING CO HOUSTON TX 29

  2. Design methods for high temperature power plant structures

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1984-01-01

    The subject is discussed under the headings: introduction (scope of paper - reviews of design methods and design criteria currently in use for both nuclear and fossil fuelled power plant; examples chosen are (a) BS 1113, representative of design codes employed for power station boiler plant; (b) ASME Code Case N47, which is being developed for high temperature nuclear reactors, especially the liquid metal fast breeder reactor); design codes for power station boilers; Code Case N47 (design in the absence of thermal shock and thermal fatigue; design against cyclic loading at high temperature; further research in support of high temperature design methods and criteria for LMFBRs); concluding remarks. (U.K.)

  3. Techno-economic design optimization of solar thermal power plants

    OpenAIRE

    Morin, G.

    2011-01-01

    A holistic view is essential in the engineering of technical systems. This thesis presents an integrative approach for designing solar thermal power plants. The methodology is based on a techno-economic plant model and a powerful optimization algorithm. Typically, contemporary design methods treat technical and economic parameters and sub-systems separately, making it difficult or even impossible to realize the full optimization potential of power plant systems. The approach presented here ov...

  4. Demonstration of near-real-time accounting at the AGNS Barnwell Plant

    International Nuclear Information System (INIS)

    Cobb, D.D.; Dayem, H.A.; Baker, A.L.

    1981-01-01

    Near-real-time nuclear materials accounting is being demonstrated in a series of experiments at the Allied-General Nuclear Services Barnwell Nuclear Fuels Plant. Each experiment consists of operating the second and third plutonium cycles continuously for 1 week using uranium solutions. Process data are collected in near-real time by the AGNS computerized nuclear materials control and accounting system, and the data are analyzed for diversion using decision analysis techniques developed and implemented by Los Alamos. Although the measurement system primarily consists of process control measurements that have not been optimized for near-real-time accounting, the results of a series of diversion tests show that diversion and unexpected losses from the process area can be detected

  5. Controlling system for an experimental demonstration plant for energy conversion using PEMFCs

    International Nuclear Information System (INIS)

    Culcer, Mihail; Iliescu, Mariana; Stefanescu, Ioan; Raceanu, Mircea; Enache, Adrian; Patularu, Laurentiu

    2006-01-01

    Full text: In the last decades of the previous century, due to global environmental problems, energy security and supply issues, many studies were conducted to investigate the uses for hydrogen energy and facilitate its penetration as an energy carrier. Subsequently, many industries worldwide began developing and producing hydrogen, hydrogen-powered vehicles, hydrogen fuel cells, and other hydrogen-based technologies. In view of the substantial long-term public and private benefits arising from hydrogen and fuel cells, the European Union and national governments throughout Europe, including the Romanian one, are working towards developing a consistent policy framework preparing the transition to a hydrogen based economy. ICIT Rm Valcea developed a research program on energy conversion using fuel cells, a project supported by the Romanian Ministry of Education and Research within the National R and D Program. An experimental demonstration pilot plant of energy conversion using PEMFCs and hydrogen producing via steam methane reforming (SMR) was achieved in order to investigate the development of small-scale SMR technologies and to allow testing and developing of specific components. The paper deals with the dedicated controlling system that provides automated data acquisition, manual or 'on line' operational control, gas management, humidification, temperature and flow controls of the pilot plant. (authors)

  6. Follow-up study of the MASSAHAKE-demonstration plant; MASSAHAKE-demonstraatiolaitoksen seuranta

    Energy Technology Data Exchange (ETDEWEB)

    Hartikainen, T. [Pohjois-Satakunnan Massahake Oy, Kankaanpaeae (Finland)

    1996-12-31

    First thinnings of high harvesting costs and low timber accumulation have often remained unharvested in Northern Satakunta due to unprofitability of harvesting. One possible solution for the problem is a harvesting chain based on partial-tree harvesting combined with the MASSAHAKE method. The pulpwood-chipping plant owned by Pohjois-Satakunnan MASSAHAKE Oy started operation in May 1995. The objective of this research is to clear-up the technical operability and profitability of the Kankaanpaeae demonstration plant, and the suitability of the products for industrial purposes. The second aim is to develop a delivery method, based on partial-tree harvesting, and the delivery organisation suitable for the conditions in Pohjois-Satakunta. The wood delivery of the MASSAHAKE is concentrated to first thinning forests. The first thinning area, given in the felling plan, located at the delivery area of MASSAHAKE, is 8870 ha/a. This corresponds to 283 000 m{sup 3} pulpwood, the total amount of biomass being 360 000 m{sup 3}. Felling is mainly carried out as labour input using conveyance-felling method. The biomass yield in typical birch first-thinning cut as partial-trees with top diameter of 4 cm is about 40 % higher than in harvesting with short-wood method. The unit costs of harvesting are about a third lower

  7. Follow-up study of the MASSAHAKE-demonstration plant; MASSAHAKE-demonstraatiolaitoksen seuranta

    Energy Technology Data Exchange (ETDEWEB)

    Hartikainen, T [Pohjois-Satakunnan Massahake Oy, Kankaanpaeae (Finland)

    1997-12-31

    First thinnings of high harvesting costs and low timber accumulation have often remained unharvested in Northern Satakunta due to unprofitability of harvesting. One possible solution for the problem is a harvesting chain based on partial-tree harvesting combined with the MASSAHAKE method. The pulpwood-chipping plant owned by Pohjois-Satakunnan MASSAHAKE Oy started operation in May 1995. The objective of this research is to clear-up the technical operability and profitability of the Kankaanpaeae demonstration plant, and the suitability of the products for industrial purposes. The second aim is to develop a delivery method, based on partial-tree harvesting, and the delivery organisation suitable for the conditions in Pohjois-Satakunta. The wood delivery of the MASSAHAKE is concentrated to first thinning forests. The first thinning area, given in the felling plan, located at the delivery area of MASSAHAKE, is 8870 ha/a. This corresponds to 283 000 m{sup 3} pulpwood, the total amount of biomass being 360 000 m{sup 3}. Felling is mainly carried out as labour input using conveyance-felling method. The biomass yield in typical birch first-thinning cut as partial-trees with top diameter of 4 cm is about 40 % higher than in harvesting with short-wood method. The unit costs of harvesting are about a third lower

  8. Synthesis Gas Demonstration Plant Program, Phase I. Commercial plant conceptual design and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-01

    This volume contains the trade-off study optimizing operating pressure (1200 psig was chosen), gas purification alternatives (Rectisol and Selexol processes were chosen). Coal preparation (wet grinding in a rod mill with trommel screen removal of oversize was recommended), air quality control (a 99.65% efficiency electrostatic precipitator and Wellman-Lord sulfur dioxide removal process were recommended), and for cooling tower optimization, a cooled water temperature of 83/sup 0/F was the optimum economic choice, with a hot water entering temperature of 118/sup 0/F. (LTN)

  9. Demonstrating compliance with WAPS 1.3 in the Hanford waste vitrification plant process

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, M.F.; Piepel, G.F.; Simpson, D.B.

    1996-03-01

    The high-level waste (HLW) vitrification plant at the Hanford Site was being designed to immobilize transuranic and high-level radioactive waste in borosilicate glass. This document describes the statistical procedure to be used in verifying compliance with requirements imposed by Section 1.3 of the Waste Acceptance Product Specifications (WAPS, USDOE 1993). WAPS 1.3 is a specification for ``product consistency,`` as measured by the Product Consistency Test (PCT, Jantzen 1992b), for each of three elements: lithium, sodium, and boron. Properties of a process batch and the resulting glass are largely determined by the composition of the feed material. Empirical models are being developed to estimate some property values, including PCT results, from data on feed composition. These models will be used in conjunction with measurements of feed composition to control the HLW vitrification process and product.

  10. Plant availability design aspects of Korean next generation reactor

    International Nuclear Information System (INIS)

    Woo Sang Lim; Ha Chung Beak

    1998-01-01

    The purpose of this paper is to describe the KNGR design concepts adopted for reducing forced outages and refueling outages, and current design changes, to assess their availability impacts compared to existing domestic nuclear power plants, and then to identify design directions for next design stage. (author)

  11. The System 80+ Standard Plant design control document. Volume 20

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains 2 technical specifications bases as part of Appendix 16 A Tech Spec Bases. They are TS B3.8 Electrical Power Technical Systems Bases and TS B3.9 Refueling Operations Bases. All 3 parts of section 17 (QA) and all 10 parts of section 18 (Human Factors) of the ADM Design and Analysis are contained in this volume. Topics covered in section 17 are: design phase QA; operations phase QA; and design phase reliability assurance. Topics covered by section 18 are: design team organization; design goals; design process; functional task analysis; control room configuration; information presentation; control and monitoring; verification and validation; and review documents

  12. Advanced nuclear plant design options to cope with external events

    International Nuclear Information System (INIS)

    2006-02-01

    With the stagnation period of nuclear power apparently coming to an end, there is a renewed interest in many Member States in the development and application of nuclear power plants (NPPs) with advanced reactors. Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR Finland for Finland and France) and more are under consideration. There is a noticeable progress in the development and demonstration of innovative high temperature gas cooled reactors, for example, in China, South Africa and Japan. The Generation IV International Forum has defined the International Near Term Deployment programme and, for a more distant perspective, six innovative nuclear energy systems have been selected and certain R and D started by several participating countries. National efforts on design and technology development for NPPs with advanced reactors, both evolutionary and innovative, are ongoing in many Member States. Advanced NPPs have an opportunity to be built at many sites around the world, with very broad siting conditions. There are special concerns that safety of these advanced reactors may be challenged by external events following new scenarios and failure modes, different from those well known for the currently operated reactors. Therefore, the engineering community identified the need to assess the proposed design configurations in relation to external scenarios at the earliest stages of the design development. It appears that an early design optimization in relation to external events is a necessary requirement to achieve safe and economical advanced nuclear power plants. Reflecting on these developments, the IAEA has planned the preparation of a report to define design options for protection from external event impacts in NPPs with evolutionary and innovative reactors. The objective of this publication is to present the state-of-the-art in design approaches for the protection of NPPs with evolutionary and innovative

  13. {sup 137}Cs sorption into bentonite from Cidadap-Tasikmalaya as buffer material for disposal demonstration plant facility at Serpong

    Energy Technology Data Exchange (ETDEWEB)

    Setiawan, B., E-mail: bravo@batan.go.id; Sriwahyuni, H., E-mail: bravo@batan.go.id; Ekaningrum, NE., E-mail: bravo@batan.go.id; Sumantry, T., E-mail: bravo@batan.go.id [Radwaste Technology Center-National Nuclear Energy Agency, PUSPIPTEK, Serpong-Tangerang 15310 (Indonesia)

    2014-03-24

    According to co-location principle, near surface disposal type the disposal demonstration plant facility will be build at Serpong nuclear area. The facility also for anticipation of future needs to provide national facility for the servicing of radwaste management of non-nuclear power plant activity in Serpong Nuclear Area. It is needs to study the material of buffer and backfill for the safety of demonstration plant facility. A local bentonite rock from Cidadap-Tasikmalaya was used as the buffer materials. Objective of experiment is to find out the specific data of sorption characteristic of Cidadap bentonite as buffer material in a radwaste disposal system. Experiments were performed in batch method, where bentonite samples were contacted with CsCl solution labeled with Cs-137 in 100 ml/g liquid:solid ratio. Initial Cs concentration was 10{sup −8} M and to study the effects of ionic strength and Cs concentration in solution, 0.1 and 1.0 M NaCl also CsCl concentration ranging 10{sup −8} - 10{sup −4} M were added in solution. As the indicator of Cs saturated in bentonite samples, Kd value was applied. Affected parameters in the experiment were contact time, effects of ionic strength and concentration of CsCl. Results showed that sorption of Cs by bentonite reached constantly after 16 days contacted, and Kd value was 10.600 ml/g. Effect of CsCl concentration on Kd value may decreased in increased in CsCl concentration. Effect of ionic strength increased according to increased in concentration of background and would effect to Kd value due to competition of Na ions and Cs in solution interacts with bentonite. By obtaining the bentonite character data as buffer material, the results could be used as the basis for making of design and the basic of performance assessment the near surface disposal facility in terms of isolation capacity of radwaste later.

  14. Full Scale Technology Demonstration of a Modern Counterrotating Unducted Fan Engine Concept. Design Report

    Science.gov (United States)

    1987-01-01

    The Unducted Fan engine (UDF trademark) concept is based on an ungeared, counterrotating, unducted, ultra-high-bypass turbofan configuration. This engine is being developed to provide a high thrust-to-weight ratio power plant with exceptional fuel efficiency for subsonic aircraft application. This report covers the design methodology and details for the major components of this engine. The design intent of the engine is to efficiently produce 25,000 pounds of static thrust while meeting life and stress requirements. The engine is required to operate at Mach numbers of 0.8 or above.

  15. Fiscal 1981 Sunshine Project research report. Development of hydrothermal power plant. Development of binary cycle power plant. Conceptual plant design; 1981 nendo nessui riyo hatsuden plant no kaihatsu / binary cycle hatsuden plant no kaihatsu seika hokokusho . Plant gainen sekkei

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-03-01

    Conceptual design was made on a 10MW class binary cycle power plant for a demonstration plant superior in reliability and profitability, under most realistic current geothermal field conditions. In the design, study was made on heat balance, main pipe system, equipment allocation, and electric system for a plant system configuration, and study was also made on preheater, evaporator, condenser, turbine and others for plant component equipment. Further study was made on optimization of mist cooling condenser, instrumentation, control, utility, and environmental measures. The following basic data were obtained through the conceptual design: plant inlet hot water temperature: 130 degrees C, plant outlet hot water temperature: 70 degrees C, hot water flow rate: 1,415t/h, working fluid: R-114, R-114 pressure in evaporator: 11.98kg/cm{sup 2} abs, R-114 evaporation temperature: 91.1 degrees C, R-114 condensation temperature: 31.0 degrees C, R- 114 flow rate: 2,265t/h, site area: 106.5m x 102.4m, building area: 48.7m x 16.8m, and building height: 13.0m. (NEDO)

  16. Conceptual design of a laser fusion power plant

    International Nuclear Information System (INIS)

    Maniscalco, J.A.; Meier, W.R.; Monsler, M.J.

    1977-01-01

    A conceptual design of a laser fusion power plant is extensively discussed. Recent advances in high gain targets are exploited in the design. A smaller blanket structure is made possible by use of a thick falling region of liquid lithium for a first wall. Major design features of the plant, reactor, and laser systems are described. A parametric analysis of performance and cost vs. design parameters is presented to show feasible design points. A more definitive follow-on conceptual design study is planned

  17. The System 80+ Standard Plant design control document. Volume 15

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains all five parts of section 12 (Radiation Protection) of the ADM Design and Analysis. Topics covered are: ALARA exposures; radiation sources; radiation protection; dose assessment; and health physics program. All six parts and appendices A and B for section 13 (Conduct of Operations) of the ADM Design and Analysis are also contained in this volume. Topics covered are: organizational structure; training program; emergency planning; review and audit; plant procedures; industrial security; sabotage protection (App 13A); and vital equipment list (App 13B)

  18. The System 80+ Standard Plant design control document. Volume 18

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains the following technical specifications of section 16 (Technical Specifications) of the ADM Design and Analysis: TS 3.3 Instrumentation; TS 3.4 Reactor Coolant System; TS 3.5 Emergency Core Cooling System; TS 3.6 Containment Systems; TS 3.7 Plant Systems; TS 3.8 Electrical Power Systems; TS 3.9 Refueling Operations; TS 4.0 Design Features; TS 5.0 Administrative Controls. Appendix 16 A Tech Spec Bases is also included. It contains the following: TS B2.0 Safety Limits Bases; TS B3.0 LCO Applicability Bases; TS B3.1 Reactivity Control Bases; TS B3.2 Power Distribution Bases

  19. Generation IV nuclear plant design strategies

    International Nuclear Information System (INIS)

    Altin, V.

    2007-01-01

    In this presentation Generation IV nuclear reactor design criteria are examined under the light of known nuclear properties of fissile and fertile nuclei. Their conflicting nature is elucidated along with the resulting inevitability of a multitude of designs. The designs selected as candidates for further development are evaluated with respect to their potential to serve the different design criteria, thereby revealing their more difficult aspects of realization and the strong research challenges lying ahead

  20. Design of demand side response model in energy internet demonstration park

    Science.gov (United States)

    Zhang, Q.; Liu, D. N.

    2017-08-01

    The implementation of demand side response can bring a lot of benefits to the power system, users and society, but there are still many problems in the actual operation. Firstly, this paper analyses the current situation and problems of demand side response. On this basis, this paper analyses the advantages of implementing demand side response in the energy Internet demonstration park. Finally, the paper designs three kinds of feasible demand side response modes in the energy Internet demonstration park.

  1. Design, construction, and operations experience with the SWSA 6 [Solid Waste Storage Area] Tumulus Disposal Demonstration

    International Nuclear Information System (INIS)

    Van Hoesen, S.D.; Van Cleve, J.E.; Wylie, A.N.; Williams, L.C.; Bolinsky, J.

    1988-01-01

    Efforts are underway at the Department of Energy facilities in Oak Ridge to improve the performance of radioactive waste disposal facilities. An engineered disposal concept demonstration involving placement of concrete encased waste on a monitored concrete pad with an earthen cover is being conducted. The design, construction, and operations experience with this project, the SWSA 6 Tumulus Disposal Demonstration, is described. 1 fig., 1 tab

  2. The System 80+ Standard Plant design control document. Volume 19

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains five technical specification bases that are part of Appendix 16 A of the ADM Design and Analysis. They are: TS B3.3 Instrumentation Bases; TS B3.4 RCS Bases; TS B3.5 ECCS Bases; TS B3.6 Containment Systems Bases; and TS B3.7 Plant Systems Bases

  3. The System 80+ Standard Plant design control document. Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the following information of the CDM: (2.8) Steam and power conversion; (2.9) Radioactive waste management; (2.10) Tech Support Center; (2.11) Initial test program; (2.12) Human factors; and sections 3, 4, and 5. Also covered in this volume are parts 1--6 of section 1 (General Plant Description) of the ADM Design and Analysis

  4. Full-Carpet Design of a Low-Boom Demonstrator Concept

    Science.gov (United States)

    Ordaz, Irian; Wintzer, Mathias; Rallabhandi, Sriram K.

    2015-01-01

    The Cart3D adjoint-based design framework is used to mitigate the undesirable o -track sonic boom properties of a demonstrator concept designed for low-boom directly under the flight path. First, the requirements of a Cart3D design mesh are determined using a high-fidelity mesh adapted to minimize the discretization error of the CFD analysis. Low-boom equivalent area targets are then generated at the under-track and one off-track azimuthal position for the baseline configuration. The under-track target is generated using a trim- feasible low-boom target generation process, ensuring that the final design is not only low-boom, but also trimmed at the specified flight condition. The o -track equivalent area target is generated by minimizing the A-weighted loudness using an efficient adjoint-based approach. The configuration outer mold line is then parameterized and optimized to match the off-body pressure distributions prescribed by the low-boom targets. The numerical optimizer uses design gradients which are calculated using the Cart3D adjoint- based design capability. Optimization constraints are placed on the geometry to satisfy structural feasibility. The low-boom properties of the final design are verified using the adaptive meshing approach. This analysis quantifies the error associated with the CFD mesh that is used for design. Finally, an alternate mesh construction and target positioning approach offering greater computational efficiency is demonstrated and verified.

  5. Comparative analysis of species-based specificity in Sr 90 and Cs 137 accumulation demonstrated by ligneous plant forest communities

    International Nuclear Information System (INIS)

    Martinovich, B.S.; Vlasov, V.K.; Sak, M.M.; Golushko, R.M.; Afmogenov, A.M.; Kirykhin, O.V.

    2004-01-01

    The authors provided field-proven study of Sr 90 and Cs 137 absorption activity demonstrated by Pinus silvestris L.; Piceae abies (L.) Roth.; Quercus rubra L.; Acer platanoides L.; Betula pendula Roth.; Tilia cordata Mill, under identical habitat conditions. The above plants were examined after 5-year growth period on radionuclide-contaminated soil. To a great extent, such parameters as radionuclide accumulation in experimental plants and accumulation activity were determined by the plants' bio-ecological properties. (Authors)

  6. Water protection in coke-plant design

    Energy Technology Data Exchange (ETDEWEB)

    G.I. Alekseev [Giprokoks, the State Institute for the Design of Coke-Industry Enterprises, Kharkov (Ukraine)

    2009-07-15

    Wastewater generation, water consumption, and water management at coke plants are considered. Measures to create runoff-free water-supply and sewer systems are discussed. Filters for water purification, corrosion inhibitors, and biocides are described. An integrated single-phase technology for the removal of phenols, thiocyanides, and ammoniacal nitrogen is outlined.

  7. A Practical Optimization Method for Designing Large PV Plants

    DEFF Research Database (Denmark)

    Kerekes, Tamas; Koutroulis, E.; Eyigun, S.

    2011-01-01

    Nowadays Photovoltaic (PV) plants have multi MW sizes, the biggest plants reaching tens of MW of capacity. Such large-scale PV plants are made up of several thousands of PV panels, each panel being in the range of 150-350W. This means that the design of a Large PV power plant is a big challenge...... and configuring such a plant should be implemented taking into consideration not only the cost of the installation, but also the Annual Energy Production, the Performance Ratio and the Levelized Cost Of Energy. In this paper, an algorithm is presented including the most important models of the PV system...

  8. System Definition and Analysis: Power Plant Design and Layout

    International Nuclear Information System (INIS)

    1996-01-01

    This is the Topical report for Task 6.0, Phase 2 of the Advanced Turbine Systems (ATS) Program. The report describes work by Westinghouse and the subcontractor, Gilbert/Commonwealth, in the fulfillment of completing Task 6.0. A conceptual design for critical and noncritical components of the gas fired combustion turbine system was completed. The conceptual design included specifications for the flange to flange gas turbine, power plant components, and balance of plant equipment. The ATS engine used in the conceptual design is an advanced 300 MW class combustion turbine incorporating many design features and technologies required to achieve ATS Program goals. Design features of power plant equipment and balance of plant equipment are described. Performance parameters for these components are explained. A site arrangement and electrical single line diagrams were drafted for the conceptual plant. ATS advanced features include design refinements in the compressor, inlet casing and scroll, combustion system, airfoil cooling, secondary flow systems, rotor and exhaust diffuser. These improved features, integrated with prudent selection of power plant and balance of plant equipment, have provided the conceptual design of a system that meets or exceeds ATS program emissions, performance, reliability-availability-maintainability, and cost goals

  9. MPD3: a useful medicinal plants database for drug designing.

    Science.gov (United States)

    Mumtaz, Arooj; Ashfaq, Usman Ali; Ul Qamar, Muhammad Tahir; Anwar, Farooq; Gulzar, Faisal; Ali, Muhammad Amjad; Saari, Nazamid; Pervez, Muhammad Tariq

    2017-06-01

    Medicinal plants are the main natural pools for the discovery and development of new drugs. In the modern era of computer-aided drug designing (CADD), there is need of prompt efforts to design and construct useful database management system that allows proper data storage, retrieval and management with user-friendly interface. An inclusive database having information about classification, activity and ready-to-dock library of medicinal plant's phytochemicals is therefore required to assist the researchers in the field of CADD. The present work was designed to merge activities of phytochemicals from medicinal plants, their targets and literature references into a single comprehensive database named as Medicinal Plants Database for Drug Designing (MPD3). The newly designed online and downloadable MPD3 contains information about more than 5000 phytochemicals from around 1000 medicinal plants with 80 different activities, more than 900 literature references and 200 plus targets. The designed database is deemed to be very useful for the researchers who are engaged in medicinal plants research, CADD and drug discovery/development with ease of operation and increased efficiency. The designed MPD3 is a comprehensive database which provides most of the information related to the medicinal plants at a single platform. MPD3 is freely available at: http://bioinform.info .

  10. The System 80+ Standard Plant design control document. Volume 23

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains part 16 References and Appendix 19 A Design Alternatives for section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Also covered is section 20 Unresolved Safety Issues of the ADM Design and Analysis. Finally sections 1--6 of the ADM Emergency Operations Guidelines are contained in this volume. Information covered in these sections include: standard post-trip actions; diagnostic actions; reactor trip recovery guideline; LOCA recovery; SG tube rupture recovery

  11. Evaluation of gaseous emissions produced in the tests on the demonstration plant for sludge drying and incineration

    International Nuclear Information System (INIS)

    Lotito, V.; Spinosa, L.; Antonacci, R.; Mininni, G.

    2001-01-01

    -current operation mode. This was depending on the excessive length of the drum designed also for counter-current mode, so that ashes melted in the first part solidify in the second part because of heat losses. In conclusion, plant operation resulted satisfactory, also under the environmental aspects, despite of the very severe test conditions, with high chlorine content and the presence of aromatic compounds and copper, which acts as catalyst in the novo-synthesis of dioxins. Incineration and co-incineration of sewage sludge and hazardous compounds were therefore demonstrated to be a safe and viable disposal option [it

  12. How-To-Do-It: Using Cauliflower to Demonstrate Plant Tissue Culture.

    Science.gov (United States)

    Haldeman, Janice H.; Ellis, Jane P.

    1988-01-01

    Presents techniques used for disinfestation of plant material, preparation of equipment and media, and laboratory procedures for tissue culture using cauliflower. Details methods for preparing solutions and plant propagation by cloning. (CW)

  13. Seismic design of nuclear power plants - an assessment

    International Nuclear Information System (INIS)

    Howard, G.E.; Ibanez, P.; Smith, C.B.

    1976-01-01

    This paper presents a review and evaluation of the design standards and the analytical and experimental methods used in the seismic design of nuclear power plants with emphasis on United States practice. Three major areas were investigated: (a) soils, siting, and seismic ground motion specification; (b) soil-structure interaction; and (c) the response of major nuclear power plant structures and components. The purpose of this review and evaluation program was to prepare an independent assessment of the state-of-the-art of the seismic design of nuclear power plants and to identify seismic analysis and design research areas meriting support by the various organizations comprising the 'nuclear power industry'. Criteria used for evaluating the relative importance of alternative research areas included the potential research impact on nuclear power plant siting, design, construction, cost, safety, licensing, and regulation. (Auth.)

  14. Seismic design for Monju FBR power plant

    International Nuclear Information System (INIS)

    1982-01-01

    This technical report introduces the basic concept on the aseismatic design of the FBR ''Monju'' power station, of which the construction in Tsuruga is planned by the Power Reactor and Nuclear Fuel Development Corp. The safety design of Monju has been performed according to ''The concept of evaluating the safety of fast breeder reactors'', and the thought concerning the aseismatic design also is written in it. According to it, ''The guide for the examination of aseismatic design regarding power reactor facilities'' should be referred to, and the classification according to the importance in aseismatic design must be made, taking the features in the design of liquid metal-cooled FBRs fully in consideration. In the aseismatic design of Monju performed according to these basic concept, the following two points were examined. In the aseismatic design of the equipment and piping, the difference of construction from LWRs such as low pressure, thin walled and high temperature construction is taken in consideration. The classification according to the aseismatic importance of the system and equipment is made on the basis of the features in the design of Monju. The classification according to aseismatic importance, the method of calculating earthquake power, the combination of loads and the allowable limit, and the aseismatic construction of the main facilities are reported. (Kako, I.)

  15. MHI-Westinghouse joint FBR tank plant design

    International Nuclear Information System (INIS)

    Arnold, W.H.; Vijuk, R.M.; Aoki, I.; Meshii, T.

    1987-01-01

    Mitsubishi Heavy Industries and Westinghouse Advanced Energy Systems Division have combined their experience and capabilities to design a tank type fast breeder reactor plant. This tank type reactor has been refined and improved during the last three years to better compete in cost, satety, and operation with alternative power plants. This Mitsubishi/Westinghouse joint design offers economic advantages due to the use of steel structures, modular construction, nitrogen cells for the intermediate loops, reactor cavity air cooling and the use of the guard vessel as the containment vessel. Inherent characteristics in the reactor design provide protection to the public and the plant investment. (author)

  16. Computer code development plant for SMART design

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  17. Computer code development plant for SMART design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  18. The System 80+ Standard Plant design control document. Volume 10

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains Appendices 6A, 6B, and 6C for section 6 (Engineered Safety Features) of the ADM Design and Analysis. Also, parts 1--5 of section 7 (Instrumentation and Control) of the ADM Design and Analysis are covered. The following information is covered in these parts: introduction; reactor protection system; ESF actuation system; system required for safe shutdown; and safety-related display instrumentation

  19. The System 80+ Standard Plant design control document. Volume 17

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 2-7 and appendix 15A for section 15 (Accident Analysis) of the ADM Design and Analysis. Topics covered in these parts are: decrease in heat removal; decrease in RCS flow rate; power distribution anomalies; increase in RCS inventory; decrease in RCS inventory; release of radioactive materials. The appendix covers radiological release models. Also contained here are five technical specifications for section 16 (Technical Specifications) of the ADM Design and Analysis. They are: TS 1.0 Use and Applications; TS 2.0 Safety Limits; TS 3.0 LCO Availability; TS 3.1 Reactivity Control; and TS 3.2 Power Distribution

  20. Design of an MgB2 race track coil for a wind generator pole demonstration

    DEFF Research Database (Denmark)

    Abrahamsen, Asger Bech; Magnusson, Niklas; Jensen, Bogi Bech

    2014-01-01

    An MgB2 race track coil intended for demonstrating a down scaled pole of a 10 MW direct drive wind turbine generator has been designed. The coil consists of 10 double pancake coils stacked into a race track coil with a cross section of 84 mm × 80 mm. The length of the straight section is 0.5 m...

  1. Design of a Super-Pixel-Based Quantum Secure Authentication Demonstrator

    NARCIS (Netherlands)

    Toebes, Chris; Tentrup, Tristan B.H.; Pinkse, Pepijn W.H.

    2017-01-01

    Quantum Secure Authentication (QSA) is a method recently developed to authenticate a multiple-scattering key [1]. Previous implementations only showed proof-of-principle setups. We present a design of a compact and robust demonstration device for Quantum Secure Authentication. The challenge and

  2. Design of equipment used for high-level waste vitrification at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Vance, R.F.; Brill, B.A.; Carl, D.E.

    1997-06-01

    The equipment as designed, started, and operated for high-level radioactive waste vitrification at the West Valley Demonstration Project in western New York State is described. Equipment for the processes of melter feed make-up, vitrification, canister handling, and off-gas treatment are included. For each item of equipment the functional requirements, process description, and hardware descriptions are presented

  3. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    Mokhov, V.A.

    2010-01-01

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  4. The Effects of Designated Pollutants on Plants

    Science.gov (United States)

    1978-11-01

    Persea americana Mill. Haas and Bacon Barley Hordeum vulgare L. CM 67 Bean Phaseolus vulgaris L. Pinto, U.I. III Briza Briza maxima L. Ornamental...Tagetes patula L. French dwarf double goldie Marigold Tagetes erecta L. American ,Senator Dirksen Petunia Petunia hybrida Vilm. White cascade Radish...Probit analysis of five plant species: citrus seedlings, lemon, orange,, grape, French marigold, American marigold. Probit scale is the probability that a

  5. An approach to nuclear plant design and modification support for Russian-designed plants in Eastern Europe

    International Nuclear Information System (INIS)

    Ioannidi, J.; Akins, M.J.

    2002-01-01

    The Western nuclear countries have embarked on numerous programs to improve the safety of the Russian-designed nuclear power plants. In Russian-designed plants in Eastern Europe, plant management is being asked for the first time to decide which safety projects to implement and is finding itself lacking in nuclear safety analytical tools and practices, funds, and experience with project management and project engineering skills and tools. Some of the major areas where assistance is needed are: 1) Defining plant weaknesses toward nuclear safety. 2) Evaluating and grading the importance to safety of proposed modification. 3) Project Planning and Scheduling using computer based scheduling software. 4) Project Finance Development and Management using well defined cash flow management techniques. 5) Contract Management and Change Control. 6) Interface Management. Each of these areas requires a significant amount of discussion to understand the issues and problems associated with them. However, this paper is limited to the Project Management areas. This paper encourages the use of a design engineering firm experienced in safety practices and associated management and technical skills to serve as the Owner's Engineer/Project Management Consultant for the program period for a Russian-designed plants located outside Russia. This approach would allow for the availability and transfer of knowledge of safety practices to plant personnel and owners engineers at nuclear plants outside Russia, improving their nuclear safety culture. The plant personnel would control plant modernizations and upgrades based upon a proven and well-defined process for detailed project definition, configuration change control, and project management. This offers the opportunity to enhance the long-term safety culture by developing plant personnel knowledgeable of the safety practices, plant design basis, developing a modification control process enabling them to control the design basis through future

  6. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    International Nuclear Information System (INIS)

    Grasso, G.; Petrovich, C.; Mattioli, D.; Artioli, C.; Sciora, P.; Gugiu, D.; Bandini, G.; Bubelis, E.; Mikityuk, K.

    2014-01-01

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MW th ) and of its demonstrator reactor (300 MW th ) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors

  7. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Petrovich, C., E-mail: carlo.petrovich@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Artioli, C., E-mail: carlo.artioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sciora, P., E-mail: pierre.sciora@cea.fr [CEA (Alternative Energies and Atomic Energy Commission), DEN, DER, 13108 St Paul lez Durance (France); Gugiu, D., E-mail: daniela.gugiu@nuclear.ro [RATEN-ICN (Institute for Nuclear Research), Cod 115400 Mioveni, Str. Campului, 1, Jud. Arges (Romania); Bandini, G., E-mail: giacomino.bandini@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Bubelis, E., E-mail: evaldas.bubelis@kit.edu [KIT (Karlsruhe Institute of Technology), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [PSI (Paul Scherrer Institute), OHSA/D11, 5232 Villigen PSI (Switzerland)

    2014-10-15

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MW{sub th}) and of its demonstrator reactor (300 MW{sub th}) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors.

  8. Engineering design and test plan for demonstrating DETOX treatment of mixed wastes

    International Nuclear Information System (INIS)

    Goldblatt, S.; Dhooge, P.

    1995-01-01

    DETOX is a cocatalyzed wet oxidation process in which the catalysts are a relatively great concentration of iron ions (typically as iron(III) chloride) in the presence of small amounts of platinum and ruthenium ions. Organic compounds are oxidized completely to carbon dioxide, water, and (if chlorinated) hydrogen chloride. The process has shown promise as a non-thermal alternative to incineration for treatment and/or volume reduction of hazardous, radioactive, and mixed wastes. Design and fabrication of a demonstration unit capable of destroying 25. Kg/hr of organic material is now in progress. This paper describes the Title 2 design of the demonstration unit, and the planned demonstration effort at Savannah River Site (SRS) and Weldon Spring Site Remedial Action Project (WSSRAP)

  9. E-ELT M5 field stabilisation unit scale 1 demonstrator design and performances evaluation

    Science.gov (United States)

    Casalta, J. M.; Barriga, J.; Ariño, J.; Mercader, J.; San Andrés, M.; Serra, J.; Kjelberg, I.; Hubin, N.; Jochum, L.; Vernet, E.; Dimmler, M.; Müller, M.

    2010-07-01

    The M5 Field stabilization Unit (M5FU) for European Extremely Large Telescope (E-ELT) is a fast correcting optical system that shall provide tip-tilt corrections for the telescope dynamic pointing errors and the effect of atmospheric tiptilt and wind disturbances. A M5FU scale 1 demonstrator (M5FU1D) is being built to assess the feasibility of the key elements (actuators, sensors, mirror, mirror interfaces) and the real-time control algorithm. The strict constraints (e.g. tip-tilt control frequency range 100Hz, 3m ellipse mirror size, mirror first Eigen frequency 300Hz, maximum tip/tilt range +/- 30 arcsec, maximum tiptilt error < 40 marcsec) have been a big challenge for developing the M5FU Conceptual Design and its scale 1 demonstrator. The paper summarises the proposed design for the final unit and demonstrator and the measured performances compared to the applicable specifications.

  10. ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets

    Energy Technology Data Exchange (ETDEWEB)

    Sorbom, B.N., E-mail: bsorbom@mit.edu; Ball, J.; Palmer, T.R.; Mangiarotti, F.J.; Sierchio, J.M.; Bonoli, P.; Kasten, C.; Sutherland, D.A.; Barnard, H.S.; Haakonsen, C.B.; Goh, J.; Sung, C.; Whyte, D.G.

    2015-11-15

    Highlights: • ARC reactor designed to have 500 MW fusion power at 3.3 m major radius. • Compact, simplified design allowed by high magnetic fields and jointed magnets. • ARC has innovative plasma physics solutions such as inboardside RF launch. • High temperature superconductors allow high magnetic fields and jointed magnets. • Liquid immersion blanket and jointed magnets greatly simplify tokamak reactor design. - Abstract: The affordable, robust, compact (ARC) reactor is the product of a conceptual design study aimed at reducing the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a ∼200–250 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Q{sub p} ≈ 13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ∼63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ∼23 T peak field on coil achievable with newly available REBCO superconductor technology. External current drive is provided by two innovative inboard RF launchers using 25 MW of lower hybrid and 13.6 MW of ion cyclotron fast wave power. The resulting efficient current drive provides a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing fluorine lithium beryllium (FLiBe) molten salt. The liquid blanket is low-risk technology and provides effective neutron moderation and shielding, excellent

  11. Methodology for Plantwide Design and Optimization of Wastewater Treatment Plants

    DEFF Research Database (Denmark)

    Maria Dragan, Johanna; Zubov, Alexandr; Sin, Gürkan

    2017-01-01

    Design of Wastewater Treatment Plants (WWTPs) is a complex engineering task which requires integration of knowledge and experience from environmental biotechnology, process engineering, process synthesis and design as well as mathematical programming. A methodology has been formulated and applied...... for the systematic analysis and development of plantwide design of WWTPs using mathematical optimization and statistical methods such as sensitivity and uncertainty analyses....

  12. Design and Evaluation of a Secure Virtual Power Plant.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Jay Tillay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    For three years, Sandia National Laboratories, Georgia Institute of Technology, and University of Illinois at Urbana-Champaign investigated a smart grid vision in which renewable-centric Virtual Power Plants (VPPs) provided ancillary services with interoperable distributed energy resources (DER). This team researched, designed, built, and evaluated real-time VPP designs incorporating DER forecasting, stochastic optimization, controls, and cyber security to construct a system capable of delivering reliable ancillary services, which have been traditionally provided by large power plants or other dedicated equipment. VPPs have become possible through an evolving landscape of state and national interconnection standards, which now require DER to include grid-support functionality and communications capabilities. This makes it possible for third party aggregators to provide a range of critical grid services such as voltage regulation, frequency regulation, and contingency reserves to grid operators. This paradigm (a) enables renewable energy, demand response, and energy storage to participate in grid operations and provide grid services, (b) improves grid reliability by providing additional operating reserves for utilities, independent system operators (ISOs), and regional transmission organization (RTOs), and (c) removes renewable energy high-penetration barriers by providing services with photovoltaics and wind resources that traditionally were the jobs of thermal generators. Therefore, it is believed VPP deployment will have far-reaching positive consequences for grid operations and may provide a robust pathway to high penetrations of renewables on US power systems. In this report, we design VPPs to provide a range of grid-support services and demonstrate one VPP which simultaneously provides bulk-system energy and ancillary reserves.

  13. The System 80+ Standard Plant design control document. Volume 1

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the DCD introduction and contains sections 1 and parts 1--7 of section 2 of the CDM. Parts 1--7 included the following: (2.1) Design of SSC; (2.2) Reactor; (2.3) RCS and connected systems; (2.4) Engineered Safety Features; (2.5) Instrumentation and Control; (2.6) Electric Power; and (2.7) Auxiliary Systems

  14. The System 80+ Standard Plant design control document. Volume 11

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers parts 6 and 7 and appendix 7A for section 7 (Instrumentation and Control) of the ADM Design and Analysis. The topics covered by these are: other systems required for safety; control systems not required by safety; and CMF evaluation of limiting faults. Parts 1--3 of section 8 (Electric Power) of the ADM are also included in this volume. Topics covered by these parts are: introduction; offsite power system; and onsite power system

  15. The System 80+ Standard Plant design control document. Volume 21

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 1--10 of section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Topics covered are: methodology; initiating event evaluation; accident sequence determination; data analysis; systems analysis; external events analysis; shutdown risk assessment; accident sequence quantification; and sensitivity analysis. Also included in this volume are Appendix 19.8A Shutdown Risk Assessment and Appendix A to Appendix 19.8A Request for Information

  16. Probabilistic risk criteria and their application to nuclear chemical plant design

    International Nuclear Information System (INIS)

    Arthur, T.; Barnes, D.S.; Brown, M.L.; Taig, A.R.; Johnston, B.D.; Hayns, M.

    1989-01-01

    A nuclear chemical plant safety strategy is presented. The use of risk criteria in design is demonstrated by reference to a particular area of the plant. This involves the application of Probabilistic Risk Assessment (PRA) techniques. Computer programs developed by the UK Atomic Energy Authority (UKAEA) at its Safety and Reliability Directorate (SRD) are used toe valuate and analyze the resultant fault trees. the magnitude of releases are estimated and individual and societal risks determined. The paper concludes that the application of PRA to a nuclear chemical plant can be structured in such a way as to allow a designer to work to quantitative risk targets

  17. Design basis reconstitution and configuration management of nuclear power plants

    International Nuclear Information System (INIS)

    Smith, P.R.

    1989-01-01

    The major design requirements of nuclear power plant components, systems, and structures are found in the plant's licensing commitments documented in the Final Safety Analysis Report and in the technical specification commitments of the plant. These specifications consider the original design and its degradation by in-service use. Before a nuclear power plant begins operation, the plant systems, structures, and organizational elements are functionally arranged to operate in a particular way. This functional arrangement is specified by the plant's design requirements and is called its configuration. The paper discusses configuration management and information management for configuration management. The management of large amounts of information and the various information systems associated with nuclear generating facilities is an ever-growing challenge for utilities. Plant operations involve a complex interrelation among data elements, especially in relation to design modifications and operational changes. Consequently, the operation of these data systems is interrelated and, as a result, redundant data items may exist. Thus, in view of the need to control and manage the plant configuration baseline, managers are striving to streamline their information management programs, which usually involves the integration of data-base systems

  18. Safety criteria for design of nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

    In Finland the general safety requirements for nuclear power plants are presented in the Council of State Decision (395/91). In this guide, safety principles which supplement the Council of State Decision and which are to be used in the design of nuclear power plants are defined

  19. Advanced nuclear power plant design with minimized use of cables

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The objective of this report is to present a nuclear power plant design with a minimum utilization of cables. The report describes the types of software and hardware that will be needed to minimize hard-wired control and instrumentation circuits and to reduce the quantity of low voltage power cables while maintaining a high availability and reliability of the plant control systems

  20. The System 80+ Standard Plant design control document. Volume 24

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains sections 7--11 of the ADM Emergency Operations Guidelines. Topics covered are: excess steam demand recovery; loss of all feedwater; loss of offsite power; station blackout recovery; and functional recovery guideline. Appendix A Severe Accident Management Guidelines and Appendix B Lower Mode Operational Guidelines are also included

  1. A design study for a medium-scale field demonstration of the viscous barrier technology

    International Nuclear Information System (INIS)

    Moridis, G.; Yen, P.; Persoff, P.; Finsterle, S.; Williams, P.; Myer, L.; Pruess, K.

    1996-09-01

    This report is the design study for a medium-scale field demonstration of Lawrence Berkeley National Laboratory's new subsurface containment technology for waste isolation using a new generation of barrier liquids. The test site is located in central California in a quarry owned by the Los Banos Gravel Company in Los Banos, California, in heterogeneous unsaturated deposits of sand, silt, and -ravel typical of many of the and DOE cleanup sites and particularly analogous to the Hanford site. The coals of the field demonstration are (a) to demonstrate the ability to create a continuous subsurface barrier isolating a medium-scale volume (30 ft long by 30 ft wide by 20 ft deep, i.e. 1/10th to 1/8th the size of a buried tank at the Hanford Reservation) in the subsurface, and (b) to demonstrate the continuity, performance, and integrity of the barrier

  2. Human factor engineering applied to nuclear power plant design

    International Nuclear Information System (INIS)

    Manrique, A.; Valdivia, J.C.

    2007-01-01

    Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. Such a plan should state: -) Activities to be performed, and -) Creation of a Human Factor Engineering team adequately qualified. The Human Factor Engineering team is an integral part of the design team and is strongly linked to the engineering organizations but simultaneously has independence to act and is free to evaluate designs and propose changes in order to enhance human behavior. TECNATOM S.A. (a Spanish company) has been a part of the Design and Human Factor Engineering Team and has collaborated in the design of an advanced Nuclear Power Plant, developing methodologies and further implementing those methodologies in the design of the plant systems through the development of the plant systems operational analysis and of the man-machine interface design. The methodologies developed are made up of the following plans: -) Human Factor Engineering implementation in the Man-Machine Interface design; -) Plant System Functional Requirement Analysis; -) Allocation of Functions to man/machine; -) Task Analysis; -) Human-System Interface design; -) Control Room Verification and -) Validation

  3. The Design and Manufacturing of Essential oil Distillation Plant for ...

    African Journals Online (AJOL)

    Choice-Academy

    The paper presents economic value of the design and manufacturing of essential oil production plant ... system with the required precision for standard quality of oil at affordable cost. Thus, the ..... still, steam injection and distribution systems,.

  4. The Design and Manufacturing of Essential oil Distillation Plant for ...

    African Journals Online (AJOL)

    The Design and Manufacturing of Essential oil Distillation Plant for rural poverty ... The adaptation of oil distillation technology for essential oil production is ... based on local resources and the first prototype has been manufactured and tested.

  5. Design verification for large reprocessing plants (Proposed procedures)

    International Nuclear Information System (INIS)

    Rolandi, G.

    1988-07-01

    In the 1990s, four large commercial reprocessing plants will progressively come into operation: If an effective and efficient safeguards system is to be applied to these large and complex plants, several important factors have to be considered. One of these factors, addressed in the present report, concerns plant design verification. Design verification provides an overall assurance on plant measurement data. To this end design verification, although limited to the safeguards aspects of the plant, must be a systematic activity, which starts during the design phase, continues during the construction phase and is particularly performed during the various steps of the plant's commissioning phase. The detailed procedures for design information verification on commercial reprocessing plants must be defined within the frame of the general provisions set forth in INFCIRC/153 for any type of safeguards related activities and specifically for design verification. The present report is intended as a preliminary contribution on a purely technical level, and focusses on the problems within the Agency. For the purpose of the present study the most complex case was assumed: i.e. a safeguards system based on conventional materials accountancy, accompanied both by special input and output verification and by some form of near-real-time accountancy involving in-process inventory taking, based on authenticated operator's measurement data. C/S measures are also foreseen, where necessary to supplement the accountancy data. A complete ''design verification'' strategy comprehends: informing the Agency of any changes in the plant system which are defined as ''safeguards relevant''; ''reverifying by the Agency upon receiving notice from the Operator on any changes, on ''design information''. 13 refs

  6. Seismic design criteria for special isotope separation plant structures

    International Nuclear Information System (INIS)

    Wrona, M.W.; Wuthrich, S.J.; Rose, D.L.; Starkey, J.

    1989-01-01

    This paper describes the seismic criteria for the design of the Special Isotope Separation (SIS) production plant. These criteria are derived from the applicable Department of Energy (DOE) orders, references and proposed standards. The SIS processing plant consistent of Load Center Building (LCB), Dye Pump Building (DPB), Laser Support Building (LSB) and Plutonium Processing Building (PPB). The facility-use category for each of the SIS building structures is identified and the applicable seismic design criteria and parameters are selected

  7. Simplified nuclear plant design for tomorrow's energy needs

    International Nuclear Information System (INIS)

    Slember, R.

    1989-09-01

    Commercial nuclear powered plants play an important role in the strategic energy plans of many countries throughout the world. Many energy planners agree that nuclear plants will have to supply an increasing amount of electrical energy in the 1990s and beyond. Just as other major industries are continually taking steps to update and improve existing products, the United States' nuclear industry has embarked on a program to simplify plant systems, shorten construction time and improve economics for new plant models. One of the models being developed by Westinghouse Electric Corporation and Burns and Roe Company is the Advanced Passive 600 MWe design which incorporates safety features that passively protect the reactor during assumed abnormal operating events. These passive safety systems utilize natural circulation/cooling for mitigating abnormal events and simplify plant design and operation. This type of system eliminates the need for costly active safety grade components, results in a reduction of ancillary equipment and assists in shortening construction time. The use of passive safety systems also permits design simplification of the auxiliary systems effectively reducing operating and maintenance requirements. Collectively, the AP600 design features result in a safe plant that addresses and alleviates the critical industry issues that developed in the 1980s. Further, the design addresses utility and regulatory requirements for safety, reliability, maintainability, operations and economics. Program results to date give confidence that the objectives of the Advanced Passive 600 design are achievable through overall plant simplification. The report will include timely results from the work being performed on the salient technical features of the design, plant construction and operation. Other required institutional changes, such as the prerequisite for a design which is complete and licensed prior to start of construction, will also be presented

  8. Toshiba integrated information system for design of nuclear power plants

    International Nuclear Information System (INIS)

    Abe, Yoko; Kawamura, Hirobumi; Sasaki, Norio; Takasaka, Kiyoshi

    1993-01-01

    TOSHIBA aims to secure safety, increase reliability and improve efficiency through the engineering for nuclear power plants and has been introducing Computer Aided Engineering (CAE). Up to the present, TOSHIBA has been developing computer systems which support each field of design and applying them to the design of nuclear power plants. The new design support system has been developed to integrate each of those systems in order to realize much greater improvement in accuracy and increase of reliability in design using state-of-the-art computer technology

  9. Support to design and construction of the PBMR plant

    International Nuclear Information System (INIS)

    Cazorla, F.; Moron, P.; Gonzalez, J. I.

    2010-01-01

    Developing the new reactor design to a licensable state for constructing a pilot plant is a tough task require specific resources, concerning knowledge and previous experience, which trespassing the pure scientific or technologic knowledge linked to the reactor conceptual design. Taking into consideration the experience derived from the collaboration between the South African company PBMR (PTY) Ltd.; the Pebble Bed Modular Reactor Designer, and Tecnatom SA, the article presents some of the aspects in which the companies or organization in charge of the design can demand external support to license and construct the pilot plants with guaranteed success. (Author)

  10. Next Generation Nuclear Plant Pre-Conceptual Design Report

    International Nuclear Information System (INIS)

    Larry Demick; Doug Vandel

    2007-01-01

    management of the project using commercial processes, share the cost of design and construction with the government, and secure a commercial nuclear operating company to operate the plant. In FY-07, Pre-Conceptual Design (PCD) work was completed by the NGNP Project at the INL with the objective of developing a framework in which the design and technology development of the NGNP could progress and to begin to develop bases for selection of the specific design and operational characteristics of NGNP. This work was completed by three contractor teams with extensive experience in HTGR technology, nuclear power applications, and hydrogen production. The teams were led by Westinghouse Electric Company, LLC; AREVA NP, Inc.; and General Atomics. The scope of work included completion of special studies to address key aspects of the NGNP (e.g., reactor type, power levels, power conversion system [PCS] and heat transfer/transport system [HTS] designs, licensing and end product disposition). The results of these special studies were applied to the development by each contractor of a recommended design for NGNP and a commercial version of the HTGR. These were then used to estimate costs and schedule for design; construction; licensing; startup and testing; operation; and deactivation, decontamination, and decommissioning (DD and D) of the NGNP and an economic assessment for an Nth of a kind (NOAK) commercial plant. A primary objective of this work is to identify research and development (R and D), data needs, and future studies required to support selection of key characteristics of and to support the design and licensing processes for the NGNP. The Preliminary Project Management Plan (PPMP) for the NGNP identifies two planning options that weigh a range of programmatic risks and approaches to mitigating risk. The two options are compliant with the EPAct, but emphasize different approaches to technology development risks, design and construction risks, and the extent of demonstration in

  11. Demonstration of a Porous Tube Hydroponic System to Control Plant Moisture and Growth

    Science.gov (United States)

    Dreschel, T. W.; Hall, C. R.; Foster, T. E.

    2003-01-01

    Accurate remote detection of plant health indicators such as moisture, plant pigment concentrations, photosynthetic flux, and other biochemicals in canopies is a major goal in plant research. Influencing factors include complex interactions between wavelength dependent absorbing and scattering features from backgrounds as well as canopy biochemical and biophysical constituents. Accurately controlling these factors in outdoor field studies is difficult. Early testing of a porous tube plant culture system has indicated that plant biomass production, biomass partitioning, and leaf moisture of plants can be controlled by precisely managing the root water potential. Managing nutrient solution chemistry can also control plant pigments, biochemical concentrations, plant biomass production, and photosynthetic rates. A test bed was developed which utilized the porous tube technology with the intent of evaluating remote sensing systems, spectral analyses procedures, gas-exchange, and fluorescence measurements for their ability to detect small differences in plant water status. Spectral analysis was able to detect small differences in the mean leaf water content between the treatments. However these small differences were not detectable in the gas-exchange or fluorescence measurements.

  12. Overview: Solar Electric Propulsion Concept Designs for SEP Technology Demonstration Mission

    Science.gov (United States)

    Mcguire, Melissa L.; Hack, Kurt J.; Manzella, David; Herman, Daniel

    2014-01-01

    JPC presentation of the Concept designs for NASA Solar Electric Propulsion Technology Demonstration mission paper. Multiple Solar Electric Propulsion Technology Demonstration Missions were developed to assess vehicle performance and estimated mission cost. Concepts ranged from a 10,000 kg spacecraft capable of delivering 4000 kg of payload to one of the Earth Moon Lagrange points in support of future human-crewed outposts to a 180 kg spacecraft capable of performing an asteroid rendezvous mission after launched to a geostationary transfer orbit as a secondary payload.

  13. Safety demonstration tests on thermal decomposition of nitrated solvent with nitric acid in nuclear fuel reprocessing plants. Contract research

    International Nuclear Information System (INIS)

    Tsukamoto, Michio; Takada, Junichi; Koike, Tadao; Watanabe, Koji; Uchiyama, Gunzou; Nishio, Gunji; Murata, Mikio

    2001-03-01

    The demonstration tests were conducted to investigate the safety of the ventilation system and integrity of the HEPA filters under the design basis accident (DBA) of the evaporator in the reprocessing plants. The tests were carried out by heating organic solvent (TBP/n- dodecane) mixed with nitric acid in a sealed vessel. It was possible to cause an explosive decomposition of TBP-complex formed by nitration of the solvent with nitric acid. The following was obtained by the analysis of the experimental results of the tests. From derivation by the experimental method, data on the maximum mass release rate and the maximum energy release rate in the explosion, as the solvent of 1 [kg] spouted out by the thermal decomposition, were obtained. They were 0.59 [kg/s] and 3240.3 [kJ/kg·s] respectively. The influence given on the cell ventilation system by this explosion was small and it was demonstrated that the safety of the HEPA filters could be secured. (author)

  14. Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Yang, T.F.; Lee, A.Y.; Ruck, G.W.; Prevenslik, T.V.; Smeltzer, G.

    1979-01-01

    The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm 2 ) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm 2 ) and for ISX-B 2 (11 kA/cm 2 ). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure

  15. Recent advances in design procedures for high temperature plant

    International Nuclear Information System (INIS)

    1988-01-01

    Thirteen papers cover several aspects of design for high temperature plant. These include design codes, computerized structural analysis and mechanical properties of materials at high temperatures. Seven papers are relevant for fast reactors and these are indexed separately. These cover shakedown design, design codes for thin shells subjected to cyclic thermal loading, the inelastic behaviour of stainless steels and creep and crack propagation in reactor structures under stresses caused by thermal cycling loading. (author)

  16. Human factors engineering in Clinch River Breeder plant design

    International Nuclear Information System (INIS)

    Planchon, H.P. Jr.; Kaushal, N.N.; Snider, J.

    1982-01-01

    The Clinch River Breeder Reactor Plant (CRBRP) Project formed a Control Room Task Force to ensure that lessons learned from the Three Mile Island accident are incorporated into the design. The charter for the Control Room Task Force was to review plant operations from the control room. The focus was on the man-machine interface to ensure that the systems' designs and operator actions meshed to properly support plant operation during normal and off-normal conditions. Specific items included for review are described. This paper describes the methodology utilized to accomplish the Task Forces' objectives and the results of the review

  17. Safeguards planning in a plant design process

    International Nuclear Information System (INIS)

    Heinrich, L.A.

    1977-01-01

    The safeguards efforts for the partitioning fuel cycle are considered. Included in the discussion are the organization of the safeguards study, the development of safeguards criteria, the expression of these criteria as requirements for facility design, and some preliminary details of the implementation of these requirements in facility and process layout

  18. Fast reactor system factors affecting reprocessing plant design

    International Nuclear Information System (INIS)

    Allardice, R.H.; Pugh, O.

    1982-01-01

    The introduction of a commercial fast reactor electricity generating system is very dependent on the availability of an efficient nuclear fuel cycle. Selection of fuel element constructional materials, the fuel element design approach and the reactor operation have a significant influence on the technical feasibility and efficiency of the reprocessing and waste management plants. Therefore the fast reactor processing plant requires liaison between many design teams -reactor, fuel design, reprocessing and waste management -often with different disciplines and conflicting objectives if taken in isolation and an optimised approach to determining several key parameters. A number of these parameters are identified and the design approach discussed in the context of the reprocessing plant. Radiological safety and its impact on design is also briefly discussed. (author)

  19. Development of Demonstration Facility Design Technology for Advanced Nuclear Fuel Cycle Process

    International Nuclear Information System (INIS)

    Cho, Il Je; You, G. S.; Choung, W. M.

    2010-04-01

    The main objective of this R and D is to develop the PRIDE (PyRoprocess Integrated inactive DEmonstration) facility for engineering-scale inactive test using fresh uranium, and to establish the design requirements of the ESPF (Engineering Scale Pyroprocess Facility) for active demonstration of the pyroprocess. Pyroprocess technology, which is applicable to GEN-IV systems as one of the fuel cycle options, is a solution of the spent fuel accumulation problems. PRIDE Facility, pyroprocess mock-up facility, is the first facility that is operated in inert atmosphere in the country. By using the facility, the functional requirements and validity of pyroprocess technology and facility related to the advanced fuel cycle can be verified with a low cost. Then, PRIDE will contribute to evaluate the technology viability, proliferation resistance and possibility of commercialization of the pyroprocess technology. The PRIDE evaluation data, such as performance evaluation data of equipment and operation experiences, will be directly utilized for the design of ESPF

  20. LMFBR plant design features for sodium spill and fire protection

    International Nuclear Information System (INIS)

    Palm, R.E.

    1982-01-01

    Design features have been developed for an LMFBR plant to protect the concrete structures from potential liquid spills and fires and prevent sodium-concrete reactions. The inclusion of these features in the plant design reduces the severity of design basis accident conditions imposed on containment and other critical plant structures. Steel liners are provided in cells containing radioactive sodium systems, and catch pans are located in non-radioactive sodium system cells. The design requirements and descriptions of each of these protective features are presented. The loading conditions, analytical approach and numerical results are also included. Design of concrete cell structures that are subject to high temperature effects from sodium spills is discussed. The structural design considers the influence of high temperature on design properties of concrete and carbon steel materials based on results of a comprehensive test program. The development of these design features and high temperature design considerations for the Clinch River Breeder Reactor Plant (CRBRP) are presented in this paper

  1. Design of the deformable mirror demonstration CubeSat (DeMi)

    Science.gov (United States)

    Douglas, Ewan S.; Allan, Gregory; Barnes, Derek; Figura, Joseph S.; Haughwout, Christian A.; Gubner, Jennifer N.; Knoedler, Alex A.; LeClair, Sarah; Murphy, Thomas J.; Skouloudis, Nikolaos; Merck, John; Opperman, Roedolph A.; Cahoy, Kerri L.

    2017-09-01

    The Deformable Mirror Demonstration Mission (DeMi) was recently selected by DARPA to demonstrate in-space operation of a wavefront sensor and Microelectromechanical system (MEMS) deformable mirror (DM) payload on a 6U CubeSat. Space telescopes designed to make high-contrast observations using internal coronagraphs for direct characterization of exoplanets require the use of high-actuator density deformable mirrors. These DMs can correct image plane aberrations and speckles caused by imperfections, thermal distortions, and diffraction in the telescope and optics that would otherwise corrupt the wavefront and allow leaking starlight to contaminate coronagraphic images. DeMi is provide on-orbit demonstration and performance characterization of a MEMS deformable mirror and closed loop wavefront sensing. The DeMi payload has two operational modes, one mode that images an internal light source and another mode which uses an external aperture to images stars. Both the internal and external modes include image plane and pupil plane wavefront sensing. The objectives of the internal measurement of the 140-actuator MEMS DM actuator displacement are characterization of the mirror performance and demonstration of closed-loop correction of aberrations in the optical path. Using the external aperture to observe stars of magnitude 2 or brighter, assuming 3-axis stability with less than 0.1 degree of attitude knowledge and jitter below 10 arcsec RMSE, per observation, DeMi will also demonstrate closed loop wavefront control on an astrophysical target. We present an updated payload design, results from simulations and laboratory optical prototyping, as well as present our design for accommodating high-voltage multichannel drive electronics for the DM on a CubeSat.

  2. Methodology for modular nuclear plant design and construction

    International Nuclear Information System (INIS)

    Lapp, C.W.; Golay, M.

    1992-01-01

    During the past decade, the rising cost of nuclear power plant construction has caused the cancellation of many projects and has forced some utilities into bankruptcy. Many factors have contributed to capital cost increases, including regulatory changes, the absence of standard designs, and low worker productivity. Low worker productivity can be attributed to the conventional building process, which is not conductive to productive labor. This study presents innovative ways to reduce the capital cost of nuclear plants through more efficient construction processes designed to increase worker productivity. A major portion of the plant capital cost is the interest paid during construction on borrowed capital. Modular fabrication could potentially reduce interest payments by compressing the construction schedule of nuclear facilities. Additional cost savings expected from modular designs arise from improved quality, productivity, and schedule control in fabrication of plant elements within a factory environment

  3. Water Treatment Pilot Plant Design Manual: Low Flow Conventional/Direct Filtration Water Treatment Plant for Drinking Water Treatment Studies

    Science.gov (United States)

    This manual highlights the project constraints and concerns, and includes detailed design calculations and system schematics. The plant is based on engineering design principles and practices, previous pilot plant design experiences, and professional experiences and may serve as ...

  4. An ergonomics action research demonstration: integrating human factors into assembly design processes.

    Science.gov (United States)

    Village, J; Greig, M; Salustri, F; Zolfaghari, S; Neumann, W P

    2014-01-01

    In action research (AR), the researcher participates 'in' the actions in an organisation, while simultaneously reflecting 'on' the actions to promote learning for both the organisation and the researchers. This paper demonstrates a longitudinal AR collaboration with an electronics manufacturing firm where the goal was to improve the organisation's ability to integrate human factors (HF) proactively into their design processes. During the three-year collaboration, all meetings, workshops, interviews and reflections were digitally recorded and qualitatively analysed to inform new 'actions'. By the end of the collaboration, HF tools with targets and sign-off by the HF specialist were integrated into several stages of the design process, and engineers were held accountable for meeting the HF targets. We conclude that the AR approach combined with targeting multiple initiatives at different stages of the design process helped the organisation find ways to integrate HF into their processes in a sustainable way. Researchers acted as a catalyst to help integrate HF into the engineering design process in a sustainable way. This paper demonstrates how an AR approach can help achieve HF integration, the benefits of using a reflective stance and one method for reporting an AR study.

  5. Sizes of secondary plant components for modularized IRIS balance of plant design

    International Nuclear Information System (INIS)

    Williamson, Martin; Townsend, Lawrence

    2003-01-01

    Herein we report on a conceptual design for a balance of plant (BOP) layout to coordinate with IRIS-like plants. The report consists of results of calculations that sizes of various BOP components. These calculations include the thermodynamic analyses and general sizing of the components in order to determine plant capability and plant layout for studies on modularity and transportability. Mathematical modeling of the BOP system involves a modified ORCENT2 code as well as standard heat transfer methods. Using typical values for PWR type plants, a general BOP design, and IRIS steam generator values, an ORCENT2 heat balance is carried out for the secondary side of the plant. Using the ORCENT2 output, standard heat transfer methods are then used to calculate system performance and component sizes. (author)

  6. GenoCAD Plant Grammar to Design Plant Expression Vectors for Promoter Analysis.

    Science.gov (United States)

    Coll, Anna; Wilson, Mandy L; Gruden, Kristina; Peccoud, Jean

    2016-01-01

    With the rapid advances in prediction tools for discovery of new promoters and their cis-elements, there is a need to improve plant expression methodologies in order to facilitate a high-throughput functional validation of these promoters in planta. The promoter-reporter analysis is an indispensible approach for characterization of plant promoters. It requires the design of complex plant expression vectors, which can be challenging. Here, we describe the use of a plant grammar implemented in GenoCAD that will allow the users to quickly design constructs for promoter analysis experiments but also for other in planta functional studies. The GenoCAD plant grammar includes a library of plant biological parts organized in structural categories to facilitate their use and management and a set of rules that guides the process of assembling these biological parts into large constructs.

  7. Creativity Support to Improve Health-and-Safety in Manu-facturing Plants: Demonstrating Everyday Creativity

    OpenAIRE

    Zachos, K.; Maiden, N.; Levis, S.

    2015-01-01

    This paper reports the development and deployment of digi-tal support for human creativity in a domain outside of the creative industries -- health-and-safety management in man-ufacturing plants. It reports applied research to extend a risk detection and resolution process at a world-class manufac-turing plant that produces tractors with creativity techniques and new digital support for the plant employees to use these techniques effectively as part of the risk detection and reso-lution proce...

  8. Chemical process and plant design bibliography 1959-1989

    International Nuclear Information System (INIS)

    Ray, M.S.

    1991-01-01

    This book is concerned specifically with chemical process in formation and plant equipment design data. It is a source for chemical engineers, students and academics involved in process and design evaluation. Over 500 chemical categories are included, from Acetaldehyde to zirconium Dioxide, with cross-referencing within the book to appropriate associated chemicals

  9. Design of equipment management information system for nuclear power plant

    International Nuclear Information System (INIS)

    Wang Chengyuan

    1996-01-01

    The author describes the ideas and practical method for need analysis, system function dividing, code design, program design and network disposition of equipment purchase management system of nuclear power plant during building, from the view of engineering investment control, schedule control and quality control

  10. Conceptual design of nuclear power plants database system

    International Nuclear Information System (INIS)

    Ishikawa, Masaaki; Izumi, Fumio; Sudoh, Takashi.

    1984-03-01

    This report is the result of the joint study on the developments of the nuclear power plants database system. The present conceptual design of the database system, which includes Japanese character processing and image processing, has been made on the data of safety design parameters mainly found in the application documents for reactor construction permit made available to the public. (author)

  11. Improved design on Qinshan 300 MWe nuclear power plant

    International Nuclear Information System (INIS)

    Shi Peihua; Cheng Wanli; Lu Rongliang

    1993-01-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented

  12. Improved design on Qinshan 300 MWe nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Peihua, Shi; Wanli, Cheng; Rongliang, Lu [Shanghai Nuclear Engineering Research and Design Inst. (China)

    1993-06-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented.

  13. Design aspects of radiation protection for nuclear power plants

    International Nuclear Information System (INIS)

    1985-01-01

    This Safety Guide deals with the provisions to be made in the design of thermal neutron reactor power plants to protect site personnel and the public from undue exposure to ionizing radiation during operational states and accident conditions. The effective radiation protection is a combination of good design, high quality construction and proper operation. The document gives guidance on how to satisfy the objectives contained in Subsection 2.2 and Section 9 of the Code of Practice on Design for Safety of Nuclear Power Plants

  14. Planting and care of fine hardwood seedlings: Designing hardwood tree plantings for wildlife

    Science.gov (United States)

    Brian J. MacGowan

    2003-01-01

    Woody plants can be of value to many wildlife species. The species of tree or shrub, or the location, size, and shape of planting can all have an impact on wildlife. The purpose of this paper is to discuss the benefits of trees and shrubs for wildlife and how to design tree and shrub plantings for wildlife. Some of the practices may conflict with other management...

  15. OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 2: Designs, Assessments, and Comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C. D.; Laybourne, P.; Shillito, K. R.

    1992-03-01

    This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability, economics, and technology development needs.

  16. Progress of independent feasibility study for modular HTGR demonstration plant to be built in China

    International Nuclear Information System (INIS)

    He Jiachen

    1989-01-01

    Many regions in China are suffering from shortage of energy as a result of the rapid growth of the national economy, for example, the growth rate of national production in 1988 reached 11.2%. A great number of coal fired plants have been built in many industrial areas. However, the difficulties relating to the transportation of coal and environmental pollution have become more and more serious. The construction of hydropower plants is limited due to uneven geographic conditions and seasons. For these reasons China needs to develop nuclear power plants. Nowadays, it has been decided, that PWR will be the main reactor type in our country, but in some districts or under some conditions modular HTGR may have distinct advantages and become an attractive option. The possible plant site description and preliminary result of economic analysis of modular HTGR type reactor are briefly discussed in this presentation

  17. The first in Poland demonstrative ORC power plant of low power output

    Energy Technology Data Exchange (ETDEWEB)

    Nowak, Wladyslaw; Borsukiewicz-Gozdur, Aleksandra; Stachel, Aleksander A. [West Pomeranian Univ. of Technology, Szczecin (Poland); Klonowicz, Wojciech; Hanausek, Pawel [Turboservice Sp. z o.o., Lodz (Poland); Klonowicz, Piotr; Magiera, Radomir [Lodz Univ. of Technology (Poland)

    2010-07-01

    A description of the power plant working according to the organic Clausius-Rankine cycle (ORC) and developed at the Department of Heat Engineering (KTC), West Pomeranian University of Technology in Szczecin, is presented. The ORC power plant is powered by the low temperature heat of hot water with the temperature of up to 100 C. The hot water heat is here converted into mechanical energy that is generated by a turbine and used to drive a centrifugal air compressor. The ORC turbine is supplied with dry, saturated vapour of the R227ea working fluid of low boiling point. The working fluid vapour is generated in a combined preheater-evaporator heat exchanger. The results of calculations and experimental measurements are presented and supplemented with conclusions derived from the ORC power plant operation. Perspective modernization of the ORC power plant scheme is also outlined. (orig.)

  18. Management of design support for nuclear plant modifications

    International Nuclear Information System (INIS)

    Doyle, F.W.

    1991-01-01

    The paper will present an overview of the Ontario Hydro organization and processes for providing design support to the operating nuclear power plants. Examples of design support for Pickering GS will be highlighted. The process is described from identification of projects through the design, procurement, construction, commissioning and in-service phases. The practices for managing engineering deliverables are discussed in the context of how these integrate into the overall change control process. The interaction of Engineering with Operations, Construction, Supply and the regulatory bodies is discussed both for major retro-fit programs and for ongoing design support to the nuclear power plants. Recent experiences during the 1990 Pickering Station Outage and during the Unit 3 fuel channel replacement program are highlighted and an integrated 5 year plan for upgrading the safety related systems for the Pickering Nuclear Power Plant is presented. (author)

  19. Planning and architectural safety considerations in designing nuclear power plants

    International Nuclear Information System (INIS)

    Konsowa, Ahmed A.

    2009-01-01

    To achieve optimum safety and to avoid possible hazards in nuclear power plants, considering architectural design fundamentals and all operating precautions is mandatory. There are some planning and architectural precautions should be considered to achieve a high quality design and construction of nuclear power plant with optimum safety. This paper highlights predicted hazards like fire, terrorism, aircraft crash attacks, adversaries, intruders, and earthquakes, proposing protective actions against these hazards that vary from preventing danger to evacuating and sheltering people in-place. For instance; using safeguards program to protect against sabotage, theft, and diversion. Also, site and building well design focusing on escape pathways, emergency exits, and evacuation zones, and the safety procedures such as; evacuation exercises and sheltering processes according to different emergency classifications. In addition, this paper mentions some important codes and regulations that control nuclear power plants design, and assessment methods that evaluate probable risks. (author)

  20. Demonstration of risk-based decision analysis in remedial alternative selection and design

    International Nuclear Information System (INIS)

    Evans, E.K.; Duffield, G.M.; Massmann, J.W.; Freeze, R.A.; Stephenson, D.E.

    1993-01-01

    This study demonstrates the use of risk-based decision analysis (Massmann and Freeze 1987a, 1987b) in the selection and design of an engineering alternative for groundwater remediation at a waste site at the Savannah River Site, a US Department of Energy facility in South Carolina. The investigation focuses on the remediation and closure of the H-Area Seepage Basins, an inactive disposal site that formerly received effluent water from a nearby production facility. A previous study by Duffield et al. (1992), which used risk-based decision analysis to screen a number of ground-water remediation alternatives under consideration for this site, indicated that the most attractive remedial option is ground-water extraction by wells coupled with surface water discharge of treated effluent. The aim of the present study is to demonstrate the iterative use of risk-based decision analysis throughout the design of a particular remedial alternative. In this study, we consider the interaction between two episodes of aquifer testing over a 6-year period and the refinement of a remedial extraction well system design. Using a three-dimensional ground-water flow model, this study employs (1) geostatistics and Monte Carlo techniques to simulate hydraulic conductivity as a stochastic process and (2) Bayesian updating and conditional simulation to investigate multiple phases of aquifer testing. In our evaluation of a remedial alternative, we compute probabilistic costs associated with the failure of an alternative to completely capture a simulated contaminant plume. The results of this study demonstrate the utility of risk-based decision analysis as a tool for improving the design of a remedial alternative through the course of phased data collection at a remedial site

  1. A demonstration of on-line plant corrosion monitoring using thin layer activation

    International Nuclear Information System (INIS)

    Asher, J.; Webb, J.W.; Wilkins, N.J.M.; Lawrence, P.F.; UKAEA Atomic Energy Research Establishment, Harwell. Materials Development Div.)

    1981-12-01

    The corrosion of a 1 inch water pipe in an evaporative cooling system has been monitored over three periods of plant operation using thin layer activation (TLA). The corrosion rate was followed at a sensitivity of about 1 μm and clearly reflected changes in plant operation. Examination of the test section after removal, both by autoradiography and metallography revealed the extent of corrosion and pitting over the active area. (author)

  2. Maintenance and repair aspects of the steam generator modules for the United States' LMFBR demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Devlin, R W

    1975-07-01

    This paper describes the main considerations relating to the field maintenance and repair of the steam generator modules for the Clinch River Breeder Reactor Plant and the development approaches being employed for some of the critical elements of these operations. In particular, the approach to plant chemical cleaning of the waterside of the modules and the approach to recovery from leaks between the water and sodium sides of the modules are discussed. (author)

  3. Maintenance and repair aspects of the steam generator modules for the United States' LMFBR demonstration plant

    International Nuclear Information System (INIS)

    Devlin, R.W.

    1975-01-01

    This paper describes the main considerations relating to the field maintenance and repair of the steam generator modules for the Clinch River Breeder Reactor Plant and the development approaches being employed for some of the critical elements of these operations. In particular, the approach to plant chemical cleaning of the waterside of the modules and the approach to recovery from leaks between the water and sodium sides of the modules are discussed. (author)

  4. Human factor engineering applied to nuclear power plant design

    International Nuclear Information System (INIS)

    Manrique, A.; Valdivia, J.C.; Jimenez, A.

    2001-01-01

    For the design and construction of new nuclear power plants as well as for maintenance and operation of the existing ones new man-machine interface designs and modifications are been produced. For these new designs Human Factor Engineering must be applied the same as for any other traditional engineering discipline. Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. Additionally, the big saving achieved by a nuclear power plant having an operating methodology which significantly decreases the risk of operating errors makes it necessary and almost vital its implementation. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. (author)

  5. 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

  6. The implications of plant design on the life-time costs for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Macphee, D.S.; Hexter, B.C.; Young, M.P.; Wilson, B.J.

    1997-01-01

    Utilising the experience gained during many years of design and project management of nuclear plant, BNFL is now approaching the final stages of the construction and commissioning of the Sellafield MOX Plant (SMP) in the UK. The paper uses the SMP project to highlight the benefits of these experiences, in particular addressing the implications of the approach to plant design on life time costs. In addition to providing BNFL with a state of the art, commercial scale MOX fuel fabrication facility, the construction of this 120 tHM/yr facility, which is currently in the advanced stages of commissioning, represents a significant demonstration of the design and project management skills of BNFL Engineering Ltd. As well as meeting the main process requirements, the plant design incorporates the highest standards of safety, together with input from the future plant operators and potential customers. As befits a commercial scale plutonium handling facility, SMP also incorporates material accountancy and security provisions that will meet all international requirements. Design, construction and commissioning of this complex and highly automated plant, has benefited from a totally integrated approach to design and documentation that considers not only project implementation but also overall lifetime costs. In addition, project management techniques, developed over many years of major project construction at Sellafield, have been utilised in order to ensure successful project implementation against a background of significant technical challenge and 'fast track' timescales. (author)

  7. Plant dynamics studies towards design of plant protection system for PFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K., E-mail: natesan@igcar.gov.in [Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kasinathan, N.; Velusamy, K.; Selvaraj, P.; Chellapandi, P. [Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Analysis of various design basis events in a fast breeder reactor towards design of plant protection system. Black-Right-Pointing-Pointer Plant dynamic modeling of a sodium cooled fast breeder reactor. Black-Right-Pointing-Pointer Selection of optimum set of plant parameters for considering best plant availability. - Abstract: Prototype fast breeder reactor (PFBR) is a 500 MWe (1250 MWt) liquid sodium cooled pool type reactor currently under construction in India. For a safe and efficient operation of the plant, it is necessary that the reactor is protected from all the transients that may occur in the plant. In order to accomplish this, adequate number of SCRAM parameters is required in the plant protection system with reliable instrumentation. For identifying the SCRAM parameters, the neutronic and thermal hydraulic responses of the plant for various possible events need to be established. Towards this, a one dimensional plant dynamics code DYANA-P has been developed with thermal hydraulic models for reactor core, hot and cold pools, intermediate heat exchangers, pipelines, steam generator, primary sodium circuits and secondary sodium circuits. The code also incorporates neutron kinetics and reactivity feedback models. By a comprehensive plant dynamics study an optimum list of SCRAM parameters and the maximum permissible response time for various instruments used for deriving them have been arrived at.

  8. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  9. Rationale, Design, and Methods for Process Evaluation in the Childhood Obesity Research Demonstration Project.

    Science.gov (United States)

    Joseph, Sitara; Stevens, Andria M; Ledoux, Tracey; O'Connor, Teresia M; O'Connor, Daniel P; Thompson, Debbe

    2015-01-01

    The cross-site process evaluation plan for the Childhood Obesity Research Demonstration (CORD) project is described here. The CORD project comprises 3 unique demonstration projects designed to integrate multi-level, multi-setting health care and public health interventions over a 4-year funding period. Three different communities in California, Massachusetts, and Texas. All CORD demonstration projects targeted 2-12-year-old children whose families are eligible for benefits under Title XXI (CHIP) or Title XIX (Medicaid). The CORD projects were developed independently and consisted of evidence-based interventions that aim to prevent childhood obesity. The interventions promote healthy behaviors in children by applying strategies in 4 key settings (primary care clinics, early care and education centers, public schools, and community institutions). The CORD process evaluation outlined 3 main outcome measures: reach, dose, and fidelity, on 2 levels (researcher to provider, and provider to participant). The plan described here provides insight into the complex nature of process evaluation for consortia of independently designed multi-level, multi-setting intervention studies. The process evaluation results will provide contextual information about intervention implementation and delivery with which to interpret other aspects of the program. Copyright © 2015 Society for Nutrition Education and Behavior. All rights reserved.

  10. Revolution of Nuclear Power Plant Design Through Digital Technology

    International Nuclear Information System (INIS)

    Zhang, L.; Shi, J.; Chen, W.

    2015-01-01

    In the digital times, digital technology has penetrated into every industry. As the highest safety requirement standard, nuclear power industry needs digital technology more to breed high quality and efficiency. Digital power plant is derived from digital design and the digitisation of power plant transfer is an inevitable trend. This paper introduces the technical solutions and features of digital nuclear power plant construction by Shanghai Nuclear Engineering Research & Design Institute, points out the key points and technical difficulties that exist in the process of construction and can serve as references for further promoting construction of digital nuclear power plant. Digital technology is still flourishing. Although many problems will be encountered in construction, it is believed that digital technology will make nuclear power industry more safe, cost-effective and efficient. (author)

  11. An electron beam flue gas treatment plant for a coal fired thermal power station. EBA demonstration plant in Chengdu thermal power station (China EBA Project)

    International Nuclear Information System (INIS)

    Doi, Yoshitaka; Nakanishi, Ikuo; Shi, Jingke

    1999-01-01

    Ebara's electron beam flue gas treatment plant was installed and is being demonstrated in Chengdu Thermal Power Station, Sichuan, China. The demonstration is proving that this plant is fully capable of meeting the target removal of sulfur dioxides from flue gas (flow rate : 300-thousand m 3 /h). Recovered by-products, namely ammonium sulfate and ammonium nitrate, from the treatment were actually tested as fertilizers, the result of which was favorable. The sale and distribution of these by-products are already underway. In May 1995, this plant was presented the certificate of authorization by China's State Power Corporation. It is noted that this was the first time a sulfur dioxide removal plant was certified as such in China. (author)

  12. Nuclear power plant system environmental design and decision methodology

    International Nuclear Information System (INIS)

    Zendehrouh, Z.; Shinozuka, M.; Schauer, F.P.

    1975-01-01

    The methodology described is concerned with a system reliability analysis by which the correlation among the level of design for the environmental and natural phenomena (earthquake, flood, tornado, etc.), reasonable practical measure of safety (such as conventional safety factor), and damage (radioactivity release) probability are established. In fact, the methodology indicates how the risk of environmental and natural hazard is combined with a specific design in order to evaluate damage probability associated with the design. This leads to the optimum design decision when combined further with the cost considerations involving the radioactivity release. This fundamental approach is essential in the design of nuclear plant structures, because, unlike the convential structures, the architectural considerations and structural analysis requirements alone cannot, by themselves, result in a balanced design in the framework of social requirements. The proposed methodology incorporates the different methods of environmental load determinations with their respective probabilistic formulations as well as detailed and advanced multi-discipline (structural, mechanical, soil, nuclear physics, biology, etc.) theoretical and empirical analysis including the effect of probabilistic nature of design variables, to establish a sound and reasonable design decision model for nuclear power plants. The information required for the analysis is also described and the areas for which further research is desirable are pointed out. Furthermore, the proposed methodology can very well be utilized to determine the requirements of standardized plants to facilitate the speed of their design and review process

  13. Cycle design flexibility for the Browns Ferry Nuclear Plant

    International Nuclear Information System (INIS)

    Beu, T.D.; Keys, T.A.; Gardner, D.R.

    1986-01-01

    Fuel cycle length flexibility is being included in the cycle designs of the Browns Ferry Nuclear Plant owned and operated by the Tennessee Valley Authority. Large end-of-cycle exposure windows are accommodated by adjusting the length of end-of-cycle coastdowns. Constraints on cycle designs are being lessened by application of innovative core loading strategies and through implementation of advanced fuel designs. Changes in design bases are evaluated relatively quickly and factored into cycle designs in order to maintain or improve performance

  14. Conceptual design of inertial confinement fusion power plant

    International Nuclear Information System (INIS)

    Mima, Kunioki; Yamanaka, Tatsuhiko; Nakai, Sadao

    1994-01-01

    Presented is the status of the conceptual design studies of inertial confinement fusion reactors. The recent achievements of the laser fusion research enable us to refine the conceptual design of the power plant. In the paper, main features of several new conceptual designs of ICF reactor; KOYO, SIRIUS-P, HYLIFE-II and so on are summarized. In particular, the target design and the reactor chamber design are described. Finally, the overview of the laser fusion reactor and the irradiation system is also described. (author)

  15. Design of XML-based plant data model

    International Nuclear Information System (INIS)

    Nair, Preetha M.; Padmini, S.; Gaur, Swati; Diwakar, M.P.

    2013-01-01

    XML has emerged as an open standard for exchanging structured data on various platforms to handle rich, nested, complex data structures. XML with its flexible tree-like data structure allows a more natural representation as compared to traditional databases. In this paper we present data model for plant data acquisition systems captured using XML technologies. Plant data acquisition systems in a typical Nuclear Power Plant consists of embedded nodes at the first tier and operator consoles at the second tier for operator operation, interaction and display of Plant parameters. This paper discusses a generic data model that was designed to capture process, network architecture, communication/interface protocol and diagnostics aspects required for a Nuclear Power Plant. (author)

  16. Plant designer's view of the operator's role in nuclear plant safety

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Porter, N.J.

    1981-01-01

    The nuclear plant operator's role supports the design assumptions and equipment with four functional tasks. He must set up th plant for predictable response to disturbances, operate the plant so as to minimize the likelihood and severity of event initiators, assist in accomplishing the safety functions, and feed back operating experiences to reinforce or redefine the safety analyses' assumptions. The latter role enhances the operator effectiveness in the former three roles. The Safety Level Concept offers a different perspective that enables the operator to view his roles in nuclear plant safety. This paper outlines the operator's role in nuclear safety and classifies his tasks using the Safety Level Concept

  17. Design, construction, and operation of the contact size reduction facility at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Frank, D.E.; Reeves, S.R.; Valenti, P.J.

    1988-05-01

    This paper describes the design, construction and initial operation of the Contact-Handled Size Reduction Facility (CSRF) at the West Valley Demonstration Project. The facility was constructed to size reduce contaminated tanks, piping, and other metallic scrap and package the scrap for disposal. In addition, the CSRF has the capability to decontaminate scrap prior to disposal. The anticipated result of decontaminating the scrap is to reduce waste classified as transuranic or low-level Class B and C to Class A or release for unrestricted use as nonradioactive equipment. 10 figs., 1 tab

  18. Design and Demonstration of Emergency Control Modes for Enhanced Engine Performance

    Science.gov (United States)

    Liu, Yuan; Litt, Jonathan S.; Guo, Ten-Huei

    2013-01-01

    A design concept is presented for developing control modes that enhance aircraft engine performance during emergency flight scenarios. The benefits of increased engine performance to overall vehicle survivability during these situations may outweigh the accompanied elevated risk of engine failure. The objective involves building control logic that can consistently increase engine performance beyond designed maximum levels based on an allowable heightened probability of failure. This concept is applied to two previously developed control modes: an overthrust mode that increases maximum engine thrust output and a faster response mode that improves thrust response to dynamic throttle commands. This paper describes the redesign of these control modes and presents simulation results demonstrating both enhanced engine performance and robust maintenance of the desired elevated risk level.

  19. Conceptual design of krypton recovery plant by porous membrane method

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Fujine, Sachio; Shimizu, Toku; Saito, Keiichiro; Ouchi, Misao

    1979-10-01

    A conceptual design of a krypton recovery plant by porous membrane method was made to study feasibility of treating fuel reprocessing off-gas. Specifications of the plant could be clarified, such as off-gas pretreatment system, first cascade system of gaseous diffusion Hertz cascade composed of two-compartment diffusers, storage system, shield and housing and operating conditions. Capital costs and operating costs of the plant were estimated for different operating conditions and cost parameters. Technical and economic feasibility of the method compares favorably with those of the cryogenic distillation or the solvent absorption method. (author)

  20. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  1. Design Provisions for Withstanding Station Blackout at Nuclear Power Plants

    International Nuclear Information System (INIS)

    2015-08-01

    International operating experience has shown that the loss of off-site power supply concurrent with a turbine trip and unavailability of the standby alternating current power system is a credible event. Lessons learned from the past and recent station blackout events, as well as the analysis of the safety margins performed as part of the ‘stress tests’ conducted on European nuclear power plants in response to the Fukushima Daiichi accident, have identified the station blackout event as a limiting case for most nuclear power plants. The magnitude 9.0 earthquake and consequential tsunami which occurred in Fukushima, Japan, in March 2011, led to a common cause failure of on-site alternating current electrical power supply systems at the Fukushima Daiichi nuclear power plant as well as the off-site power grid. In addition, the resultant flooding caused the loss of direct current power supply, which further exacerbated an already critical situation at the plant. The loss of electrical power resulted in the meltdown of the core in three reactors on the site and severely restricted heat removal from the spent fuel pools for an extended period of time. The plant was left without essential instrumentation and controls, and this made accident management very challenging for the plant operators. The operators attempted to bring and maintain the reactors in a safe state without information on the vital plant parameters until the power supply was eventually restored after several days. Although the Fukushima Daiichi accident progressed well beyond the expected consequences of a station blackout, which is the complete loss of all alternating current power supplies, many of the lessons learned from the accident are valid. A failure of the plant power supply system such as the one that occurred at Fukushima Daiichi represents a design extension condition that requires management with predesigned contingency planning and operator training. The extended loss of all power at a

  2. Design and analysis of heat recovery system in bioprocess plant

    International Nuclear Information System (INIS)

    Anastasovski, Aleksandar; Rašković, Predrag; Guzović, Zvonimir

    2015-01-01

    Highlights: • Heat integration of a bioprocess plant is studied. • Bioprocess plant produces yeast and ethyl-alcohol. • The design of a heat recovery system is performed by batch pinch analysis. • Direct and indirect heat integration approaches are used in process design. • The heat recovery system without a heat storage opportunity is more profitable. - Abstract: The paper deals with the heat integration of a bioprocess plant which produces yeast and ethyl-alcohol. The referent plant is considered to be a multiproduct batch plant which operates in a semi-continuous mode. The design of a heat recovery system is performed by batch pinch analysis and by the use of the Time slice model. The results obtained by direct and indirect heat integration approaches are presented in the form of cost-optimal heat exchanger networks and evaluated by different thermodynamic and economic indicators. They signify that the heat recovery system without a heat storage opportunity can be considered to be a more profitable solution for the energy efficiency increase in a plant

  3. Design option of heat exchanger for the next generation nuclear plant - HTR2008-58175

    International Nuclear Information System (INIS)

    Oh, C. H.; Kim, E. S.

    2008-01-01

    The Next Generation Nuclear Plant (NGNP), a very High temperature Gas-Cooled Reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit (PCU) for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger is very important. This paper will include analysis of one stage versus two stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical coil heat exchanger, and shell/tube heat exchanger. (authors)

  4. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  5. Initial Development and Pilot Study Design of Interactive Lecture Demonstrations for ASTRO 101

    Science.gov (United States)

    Schwortz, Andria C.; French, D. A; Gutierrez, Joseph V; Sanchez, Richard L; Slater, Timothy F.; Tatge, Coty

    2014-06-01

    Interactive lecture demonstrations (ILDs) have repeatedly shown to be effective tools for improving student achievement in the context of learning physics. As a first step toward systematic development of interactive lecture demonstrations in ASTRO 101, the introductory astronomy survey course, a systematic review of education research, describing educational computer simulations (ECSs) reveals that initial development requires a targeted study of how ASTRO 101 students respond to ECSs in the non-science majoring undergraduate lecture setting. In this project we have adopted the process by which ILDs were designed, pilot-tested, and successfully implemented in the context of physics teaching (Sokoloff & Thornton, 1997; Sokoloff & Thornton, 2004). We have designed the initial pilot-test set of ASTRO 101 ILD instructional materials relying heavily on ECSs. Both an instructor’s manual and a preliminary classroom-ready student workbook have been developed, and we are implementing a pilot study to explore their effectiveness in communicating scientific content, and the extent to which they might enhance students’ knowledge of and perception about astronomy and science in general. The study design uses a pre-/post-test quasi-experimental study design measuring students’ normalized gain scores, calculated as per Hake (1998) and Prather (2009), using a slightly modified version of S. Slater’s (2011) Test Of Astronomy STandards TOAST combined with other instruments. The results of this initial study will guide the iterative development of ASTRO 101 ILDs that are intended to both be effective at enhancing student achievement and easy for instructors to successfully implement.

  6. Development of demonstration facility design technology for advanced nuclear fuel cycle process

    International Nuclear Information System (INIS)

    Cho, Il Je; You, G. S.; Choung, W. M.; Lee, E. P.; Hong, D. H.; Lee, W. K.; Ku, J. H.; Moon, S. I.; Kwon, K. C.; Lee, K. I. and other

    2012-04-01

    PRIDE Facility, pyroprocess mock-up facility, is the first facility that is operated in inert atmosphere in the country. By using the facility, the functional requirements and validity of pyroprocess technology and facility related to the advanced fuel cycle can be verified with a low cost. Then, PRIDE will contribute to evaluate the technology viability, proliferation resistance and possibility of commercialization of the pyroprocess technology. It is essential to develop design technologies for the advanced nuclear fuel cycle demonstration facilities and complete the detailed design of PRIDE facility with capabilities of the stringent inert atmosphere control, fully remote operation which are necessary to develop the high-temperature molten salts technology. For these, it is necessary to design the essential equipment of large scale inert cell structure and the control system to maintain the inert atmosphere, and evaluate the safety. To construct the hot cell system which is appropriate for pyroprocess, some design technologies should be developed, which include safety evaluation for effective operation and maintenance, radiation safety analysis for hot cell, structural analysis, environmental evaluation, HVAC systems and electric equipment

  7. Seismic considerations in the design of atomic power plants

    International Nuclear Information System (INIS)

    Arya, A.S.; Chandrasekaran, A.R.; Thakkar, S.K.

    1975-01-01

    A seismic design is one of the most important factors for the safety of nuclear power plants constructed in seismic areas. The various considerations in the design of atomic power plant structures and components to achieve high degree (near absolute) of safety during future probable earthquakes is described as follows: (a) determination of design earthquake parameters for SSE and OBE (b) fixing time history accelerograms and acceleration response spectra (c) mathematical modelling of the reactor building considering soil-structure interaction (d) deciding allowable stresses, damping factors and serviceability limits like drift, displacements and crack widths (e) tests for determining stiffness and damping characteristics of components in-situ before commissioning of plant. The main questions that arise under various items requiring further research investigations or development work are pointed out for discussion. (author)

  8. Computer aided design of piping for a radiochemical plant

    Energy Technology Data Exchange (ETDEWEB)

    Selvaraj, P G; Chandrasekhar, A; Chandrasekar, A V [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Raju, R P; Mahudeeswaran, K V; Kumar, S V [Reprocessing Group, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    In a radiochemical plant such as reprocessing plants, process equipment, storage tanks, liquid transfer systems and the associated pipe lines etc. are housed in series of concrete cells. Availability of limited cell space/volume, provision of various modes of liquid transfers with associated redundancies and instrumentation lines with standby alternatives increase the overall piping density. Designing such high density piping layout without interference is quite complex and needs lot of human efforts. This paper briefly describes development of computer codes for the entire scheme of design, drafting and fabrication of piping for nuclear fuel reprocessing plant. The general organisation of various programs, their functions, the complete sequence of the scheme and the flow of data are presented. High degree of reliability of each routine, considerable error checking facilities, marking legends on the drawings, provision for scaling in drafting and accuracy to the extent of one mm in layout design are some of the important features of this scheme. (author). 1 fig.

  9. Control room design and human engineering in power plants

    International Nuclear Information System (INIS)

    Herbst, L.; Hinz, W.

    1982-01-01

    The concept for modern plant control rooms is primary influenced by: The automation of protection, binary control and closed loop control functions; organization employing functional areas; computer based information processing; human engineered design. Automation reduces the human work load. Employment of functional areas permits optimization of operational sequences. Computer based information processing makes it possible to output information in accordance with operating requirements. Design based on human engineering principles assures the quality of the interaction between the operator and the equipment. The degree to which these conceptional features play a role in design of power plant control rooms depends on the unit rating, the mode of operation and on the requirements respecting safety and availability of the plant. (orig.)

  10. ELMO Bumpy Torus Reactor and power plant: conceptual design study

    International Nuclear Information System (INIS)

    Bathke, C.G.; Dudziak, D.J.; Krakowski, R.A.

    1981-08-01

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based on these results, the direction for future improvement on an already attractive reactor design is identified

  11. Optimal design of condenser volume in nuclear power plant

    International Nuclear Information System (INIS)

    Zheng Jing; Yan Changqi; Wang Jianjun

    2011-01-01

    The condenser is an important component in the nuclear power plant,whose dimension will influence the economy and the arrangement of the nuclear power plant.In this paper, the calculation model was established according to the design experience. The corresponding codes were also developed. The sensitivity of design parameters which influence the condenser Janume was analyzed. The present optimal design of the condenser, aiming at the volume minimization, was carried out with the self-developed complex-genetic algorithm. The results show that the reference condenser design is far from the best scheme. In addition, the results also verify the feasibility of the complex-genetic algorithm. Furthermore, the results of this paper can provide reference for the design of the condenser. (authors)

  12. Technical guidelines for aseismic design of nuclear power plants

    International Nuclear Information System (INIS)

    Park, Y.J.; Hofmayer, C.H.

    1994-06-01

    This document is a translation, in its entirety, of the Japan Electric Association (JEA) publication entitled open-quotes Technical Guidelines for Aseismic Design of Nuclear Power Plants - JEAG 4601-1987.close quotes This guideline describes in detail the aseismic design techniques used in Japan for nuclear power plants. It contains chapters dealing with: (a)the selection of earthquake ground motions for a site, (b) the investigation of foundation and bedrock conditions, (c) the evaluation of ground stability and the effects of ground movement on buried piping and structures, (d) the analysis and design of structures, and (e) the analysis and design of equipment and distribution systems (piping, electrical raceways, instrumentation, tubing and HVAC duct). The guideline also includes appendices which summarize data, information and references related to aseismic design technology

  13. An estimate and evaluation of design error effects on nuclear power plant design adequacy

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1984-01-01

    An area of considerable concern in evaluating Design Control Quality Assurance procedures applied to design and analysis of nuclear power plant is the level of design error expected or encountered. There is very little published data 1 on the level of error typically found in nuclear power plant design calculations and even less on the impact such errors would be expected to have on overall design adequacy of the plant. This paper is concerned with design error associated with civil and mechanical structural design and analysis found in calculations which form part of the Design or Stress reports. These reports are meant to document the design basis and adequacy of the plant. The estimates contained in this paper are based on the personal experiences of the author. In Table 1 is a partial listing of the design docummentation review performed by the author on which the observations contained in this paper are based. In the preparation of any design calculations, it is a utopian dream to presume such calculations can be made error free. The intent of this paper is to define error levels which might be expected in a competent engineering organizations employing currently technically qualified engineers and accepted methods of Design Control. In addition, the effects of these errors on the probability of failure to meet applicable design code requirements also are estimated

  14. Westinghouse AP1000 advanced passive plant: design features and benefits

    International Nuclear Information System (INIS)

    Walls, S.J.; Cummins, W.E.

    2003-01-01

    The Westinghouse AP1000 Program is aimed at implementing the AP1000 plant to provide a further major improvement in plant economics while maintaining the passive safety advantages established by the AP600. An objective is to retain to the maximum extent possible the plant design of the AP600 so as to retain the licensing basis, cost estimate, construction schedule, modularization scheme, and the detailed design from the AP600 program. Westinghouse and the US Nuclear Regulatory Commission staff have embarked on a program to complete Design Certification for the AP1000 by 2004. A pre-certification review phase was completed in March 2002 and was successful in establishing the applicability of the AP600 test program and AP600 safety analysis codes to the AP1000 Design Certification. On March 28, 2002, Westinghouse submitted to US NRC the AP1000 Design Control Document and Probabilistic Risk Assessment, thereby initiating the formal design certification review process. The results presented in these documents verify the safety performance of the API 000 and conformance with US NRC licensing requirements. Plans are being developed for implementation of a series of AP1000 plants in the US. Key factors in this planning are the economics of AP1000, and the associated business model for licensing, constructing and operating these new plants. Similarly plans are being developed to get the AP1000 design reviewed for use in the UK. Part of this planning has been to examine the AP1000 design relative to anticipated UK safety and licensing issues. (author)

  15. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz. This document, Volume IV, provides Appendix 8.B, Laboratory Investigations of Dynamic Properties of Reference Sites

  16. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz

  17. Upgrading of seismic design of nuclear power plant building

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi [Tokyo Univ. (Japan). Faculty of Engineering; Kitada, Yoshio

    1997-03-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan`s effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  18. Upgrading of seismic design of nuclear power plant building

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Kitada, Yoshio.

    1997-01-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan's effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  19. Engineering Design of ITER Prototype Fast Plant System Controller

    Science.gov (United States)

    Goncalves, B.; Sousa, J.; Carvalho, B.; Rodrigues, A. P.; Correia, M.; Batista, A.; Vega, J.; Ruiz, M.; Lopez, J. M.; Rojo, R. Castro; Wallander, A.; Utzel, N.; Neto, A.; Alves, D.; Valcarcel, D.

    2011-08-01

    The ITER control, data access and communication (CODAC) design team identified the need for two types of plant systems. A slow control plant system is based on industrial automation technology with maximum sampling rates below 100 Hz, and a fast control plant system is based on embedded technology with higher sampling rates and more stringent real-time requirements than that required for slow controllers. The latter is applicable to diagnostics and plant systems in closed-control loops whose cycle times are below 1 ms. Fast controllers will be dedicated industrial controllers with the ability to supervise other fast and/or slow controllers, interface to actuators and sensors and, if necessary, high performance networks. Two prototypes of a fast plant system controller specialized for data acquisition and constrained by ITER technological choices are being built using two different form factors. This prototyping activity contributes to the Plant Control Design Handbook effort of standardization, specifically regarding fast controller characteristics. Envisaging a general purpose fast controller design, diagnostic use cases with specific requirements were analyzed and will be presented along with the interface with CODAC and sensors. The requirements and constraints that real-time plasma control imposes on the design were also taken into consideration. Functional specifications and technology neutral architecture, together with its implications on the engineering design, were considered. The detailed engineering design compliant with ITER standards was performed and will be discussed in detail. Emphasis will be given to the integration of the controller in the standard CODAC environment. Requirements for the EPICS IOC providing the interface to the outside world, the prototype decisions on form factor, real-time operating system, and high-performance networks will also be discussed, as well as the requirements for data streaming to CODAC for visualization and

  20. Sample sequencing of vascular plants demonstrates widespread conservation and divergence of microRNAs.

    Science.gov (United States)

    Chávez Montes, Ricardo A; de Fátima Rosas-Cárdenas, Flor; De Paoli, Emanuele; Accerbi, Monica; Rymarquis, Linda A; Mahalingam, Gayathri; Marsch-Martínez, Nayelli; Meyers, Blake C; Green, Pamela J; de Folter, Stefan

    2014-04-23

    Small RNAs are pivotal regulators of gene expression that guide transcriptional and post-transcriptional silencing mechanisms in eukaryotes, including plants. Here we report a comprehensive atlas of sRNA and miRNA from 3 species of algae and 31 representative species across vascular plants, including non-model plants. We sequence and quantify sRNAs from 99 different tissues or treatments across species, resulting in a data set of over 132 million distinct sequences. Using miRBase mature sequences as a reference, we identify the miRNA sequences present in these libraries. We apply diverse profiling methods to examine critical sRNA and miRNA features, such as size distribution, tissue-specific regulation and sequence conservation between species, as well as to predict putative new miRNA sequences. We also develop database resources, computational analysis tools and a dedicated website, http://smallrna.udel.edu/. This study provides new insights on plant sRNAs and miRNAs, and a foundation for future studies.

  1. Impingement and entrainment of fishes at the Savannah River Plant: an NPDES 316b demonstration

    International Nuclear Information System (INIS)

    McFarlane, R.W.; Frietsche, R.F.; Miracle, R.D.

    1978-02-01

    Environmental impacts of the Savannah River Plant's withdrawal of Savannah River water include impingement of juvenile and adult fish on trash removal screens, and entrainment of planktonic fish eggs and larval fish into the pumping system. The Savannah River Plant (SRP) has the capacity to pump 3.6 million cubic meters of water per day--25% of the minimal river discharge--for cooling and other purposes. Present removal is 7% of the actual river discharge. In the river and intake canals reside sixty-nine species of fishes. The species composition of the resident fish community of the intake canals is similar to the species composition in the river, but different in relative species abundance. The dominant sunfishes tend to reside in the canals for long periods and seldom go from canal to canal. The fish impingement rate at the plant ranks very low in comparison with electric power plants on inland waters. Thirty-five species of fishes were impinged during 1977. The average impingement rate of 7.3 fish per day extrapolates to 2,680 fish per year. No single species comprised more than 10% of the sample. The most commonly impinged species were bluespotted sunfish, warmouth, channel catfish, and yellow perch. The relative abundance of those species impinged deviates from their relative abundance in the canal fish population

  2. The Relation of Design Parameters, Plant Capacity and Processing Costs in Cobalt-60 Sterilization Plants

    International Nuclear Information System (INIS)

    Brown, M.G.

    1967-01-01

    The paper describes the main features of three basic types of cobalt-60 sterilization plants which have been designed to provide a complete range of capacities for radiosterilization of medical products. The smallest plant has a capacity of up to 50 000 cubic feet of medical products a year, the intermediate size plant has a capacity of up to 500 000 cubic feet a year, and the largest plant has a capacity in excess of 1000 000 cubic feet a year. The relations between capital costs, rate of production, efficiency and unit processing costs for each type of plant are discussed. The method of selecting the best type of plant for a particular need will also be outlined. (author)

  3. Pre-project study on a demonstration plant for seawater desalination using a nuclear heating reactor in Morocco

    International Nuclear Information System (INIS)

    Achour, M.

    2000-01-01

    This paper gives in the first part detailed information on the pre-project study on a demonstration plant for seawater desalination using heating reactor implemented by both Moroccan and Chinese sides. The main findings of the pre-project study are given in the second part. (author)

  4. Recovery of mercury and other metals from used dry battery cells -the CJC demonstration plant in Hokkaido, Japan

    Energy Technology Data Exchange (ETDEWEB)

    Hirayama, Naomichi; Gotoh, Sukehiro; Yajima, Takenori

    1987-01-01

    The present paper deals with a project associated with the Clean Japan Center Demonstration plant, which was financially supported by the National Government and constructed recently at Itomuka, Hokkaido, for the purpose of the proper disposal of and resource recovery from mainly used dry battery cells or wastes containing mercury (Hg) and other hazardous substances. The process details are also given.

  5. A case study of radial jetting technology for enhancing geothermal energy systems at Klaipeda geothermal demonstration plant

    NARCIS (Netherlands)

    Nair, R.; Peters, E.; Sliaupa, S.; Valickas, R.; Petrauskas, S.

    2017-01-01

    In 1996 a geothermal energy project was initiated at Klaipėda, Lithuania, to demonstrate the feasibility of using low enthalpy geothermal water as a renewable energy resource in district heating systems. The Klaipėda geothermal plant is situated within the West Lithuanian geothermal anomaly with a

  6. Control room systems design for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This publication provides a resource for those who are involved in researching, managing, conceptualizing, designing, manufacturing or backfitting power plant control room systems. It will also be useful to those responsible for performing reviews or evaluations of the design and facilities associated with existing power plant control room systems. The ultimate worth of the publication, however, will depend upon how well it can support its users. Readers are invited to provide comments and observations to the IAEA, Division of Nuclear Power. If appropriate, the report will subsequently be re-issued, taking such feedback into account. Refs, figs and tabs.

  7. Control room systems design for nuclear power plants

    International Nuclear Information System (INIS)

    1995-07-01

    This publication provides a resource for those who are involved in researching, managing, conceptualizing, designing, manufacturing or backfitting power plant control room systems. It will also be useful to those responsible for performing reviews or evaluations of the design and facilities associated with existing power plant control room systems. The ultimate worth of the publication, however, will depend upon how well it can support its users. Readers are invited to provide comments and observations to the IAEA, Division of Nuclear Power. If appropriate, the report will subsequently be re-issued, taking such feedback into account. Refs, figs and tabs

  8. Control room design and human engineering in power plants

    International Nuclear Information System (INIS)

    Herbst, L.; Hinz, W.

    1981-01-01

    Automation reduces the human work load. Employment of functional areas permits optimization of operational sequences. Computer based information processing makes it possible to output information in accordance with operating requirements. Design based on human engineering principles assures the quality of the interaction between the operator and the equipment. The degree to which these conceptional features play a role in design of power plant control rooms depends on the unit rating, the mode of operation and on the requirements respecting safety and availability of the plant. (orig./RW)

  9. Design of control rooms and ergonomics in power plants

    International Nuclear Information System (INIS)

    Herbst, L.; Hinz, W.

    1981-01-01

    Modern power plant control rooms are characterized by automation of protection and control functions, subdivision according to functions, computer-aided information processing, and ergonomic design. Automation relieves the personnel of stress. Subdivision according to functions permits optimized procedures. Computer-aided information processing results in variable information output tailored to the actual needs. Ergonomic design assures qualified man-machine interaction. Of course, these characteristics will vary between power plants in dependence of unit power, mode of operation, and safety and availability requirements. (orig.) [de

  10. The current status of research and development concerning steam generator acoustic leak detection for the demonstration FBR plant

    International Nuclear Information System (INIS)

    Higuchi, Masahisa

    1990-01-01

    The Japan Atomic Power Co. (JAPC) started the research and development into Acoustic Leak Detection for the Demonstration FBR (D-FBR) plant in 1989. Acoustic Leak Detection is expected as a water leak detection system in the Steam Generator for the first D-FBR plant. JAPC is presently analyzing data on Acoustic Leak Detection in order to form some basic concepts and basic specifications about leak detection. Both low frequency types and high frequency types are selected as candidates for Acoustic Leak Detection. After a review of both types, either one will be selected for the D-FBT plant. A detailed Research and Development plan on Acoustic Leak Detection, which should be carried out prior to starting the construction of the D-FBR plant, is under review. (author). 3 figs, 2 tabs

  11. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  12. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  13. Ecohydrological Design of small hydropower plants: assessment of environmental flows

    International Nuclear Information System (INIS)

    Diez H, Juan M; Olmeda S, Sergio

    2008-01-01

    The small hydroelectric energy has contributed substantially to the progress of the life standards in the world by means of a technology with a relatively low environmental impact. The Small Hydroelectric Plant (SHP) is the convenient type of hydro energetic facility for Colombia, reason why it predictably will play a capital role in the exploitation of its excellent potential during this decade. The regulatory guidelines for the SHP re licensing enforce hydrological design criteria that incorporate the ecological requirements of discharge in the fluvial reaches affected by flow derivations. The Instream Flow (IF) should be considered as an ecological restriction to the hydroelectric use, which can be satisfactorily set with reliable approaches. This work shows the basic typology of IF assessment methods, describing the reputed Instream Flow Incremental Methodology (IFIM) with a case study of SHP re licensing. It also demonstrates the possibility of harmonize the hydroelectric operation with the maintenance of an acceptable ecological condition, preserving some IF that can be assessed with the IFIM analytic frame. The new eco hydrological methods for IF setting facilitate SHP that are profitable and defensible in environmental terms.

  14. Design and Performance of the NASA SCEPTOR Distributed Electric Propulsion Flight Demonstrator

    Science.gov (United States)

    Borer, Nicholas K.; Patterson, Michael D.; Viken, Jeffrey K.; Moore, Mark D.; Clarke, Sean; Redifer, Matthew E.; Christie, Robert J.; Stoll, Alex M.; Dubois, Arthur; Bevirt, JoeBen; hide

    2016-01-01

    Distributed Electric Propulsion (DEP) technology uses multiple propulsors driven by electric motors distributed about the airframe to yield beneficial aerodynamic-propulsion interaction. The NASA SCEPTOR flight demonstration project will retrofit an existing internal combustion engine-powered light aircraft with two types of DEP: small "high-lift" propellers distributed along the leading edge of the wing which accelerate the flow over the wing at low speeds, and larger cruise propellers co-located with each wingtip for primary propulsive power. The updated high-lift system enables a 2.5x reduction in wing area as compared to the original aircraft, reducing drag at cruise and shifting the velocity for maximum lift-to-drag ratio to a higher speed, while maintaining low-speed performance. The wingtip-mounted cruise propellers interact with the wingtip vortex, enabling a further efficiency increase that can reduce propulsive power by 10%. A tradespace exploration approach is developed that enables rapid identification of salient trades, and subsequent creation of SCEPTOR demonstrator geometries. These candidates were scrutinized by subject matter experts to identify design preferences that were not modeled during configuration exploration. This exploration and design approach is used to create an aircraft that consumes an estimated 4.8x less energy at the selected cruise point when compared to the original aircraft.

  15. Demonstration of a software design and statistical analysis methodology with application to patient outcomes data sets.

    Science.gov (United States)

    Mayo, Charles; Conners, Steve; Warren, Christopher; Miller, Robert; Court, Laurence; Popple, Richard

    2013-11-01

    With emergence of clinical outcomes databases as tools utilized routinely within institutions, comes need for software tools to support automated statistical analysis of these large data sets and intrainstitutional exchange from independent federated databases to support data pooling. In this paper, the authors present a design approach and analysis methodology that addresses both issues. A software application was constructed to automate analysis of patient outcomes data using a wide range of statistical metrics, by combining use of C#.Net and R code. The accuracy and speed of the code was evaluated using benchmark data sets. The approach provides data needed to evaluate combinations of statistical measurements for ability to identify patterns of interest in the data. Through application of the tools to a benchmark data set for dose-response threshold and to SBRT lung data sets, an algorithm was developed that uses receiver operator characteristic curves to identify a threshold value and combines use of contingency tables, Fisher exact tests, Welch t-tests, and Kolmogorov-Smirnov tests to filter the large data set to identify values demonstrating dose-response. Kullback-Leibler divergences were used to provide additional confirmation. The work demonstrates the viability of the design approach and the software tool for analysis of large data sets.

  16. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  17. International cooperation: forum and consortium for the system design and the ISER-PIUS demonstration project

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1987-01-01

    It is a wide recognition that the light water reactor (LWR) era will last long into the next century and that need is felt for establishing its foundation to be shared by all the countries concerned including now developing countries. From this perspective, the paper provides a justification account for the joint development of ISER-PIUS as a cheap and safe LWR for international use, if so allowed, with the auspices of the international nuclear agency; IAEA. Incentives for developing ISER-PIUS ranges from 'safety and nuclear economy improvements' in developed countries to 'safe and economical SMPR (small and medium sized power reactor introduction' in developing countries. The present report presents some views and comments concerning the four important issues emphasized by IAEA Director General Dr. Hans Blix: the urgency of the need to change the energy supply system in the country concerned, establishment of a suitable organizational structure, provision of qualified manpower, and required level of industrial support. An experimental and-or demonstration reactor will be a prerequisite in order both to demonstrate once for all that there is a unique reactor which can accomodate even the tests for the conventional DBA (design bases accidents) and to collect optimized design data for stable operation and maintenance of such a new reactor. (Nogami, K.)

  18. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    International Nuclear Information System (INIS)

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-01-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  19. Three-dimensional computer aided design system for plant layout

    International Nuclear Information System (INIS)

    Yoshinaga, Toshiaki; Kiguchi, Takashi; Tokumasu, Shinji; Kumamoto, Kenjiro.

    1986-01-01

    The CAD system for three-dimensional plant layout planning, with which the layout of pipings, cable trays, air conditioning ducts and so on in nuclear power plants can be planned and designed effectively in a short period is reported. This system comprises the automatic routing system by storing the rich experience and know-how of designers in a computer as the knowledge, and deciding the layout automatically following the predetermined sequence by using these, the interactive layout system for reviewing the routing results from higher level and modifying to the optimum layout, the layout evaluation system for synthetically evaluating the layout from the viewpoint of the operability such as checkup and maintenance, and the data base system which enables these effective planning and design. In this report, the total constitution of this system and the technical features and effects of the individual subsystems are outlined. In this CAD system for three-dimensional plant layout planning, knowledge engineering, CAD/CAM, computer graphics and other latest technology were introduced, accordingly by applying this system to plant design, the design can be performed quickly, various case studies can be carried out at planning stage, and systematic and optimum layout planning becomes possible. (Kako, I.)

  20. Design basis programs and improvements in plant operation

    International Nuclear Information System (INIS)

    Metcalf, M.F.

    1991-01-01

    Public Service Electric and Gas (PSE and G) Company operates three commercial nuclear power plants in southern New Jersey. The three plants are of different designs and vintages (two pressurized water reactors licensed in 1976 and 1980 and one boiling water reactor licensed in 1986). As the industry recognized the need to develop design basis programs, PSE and G also realized the need after a voluntary 52-day shutdown of one unit because of electrical design basis problems. In its drive to be a premier electric utility, PSE and G has been aggressively active in developing design basis documents (DBDs) with supporting projects and refined uses to obtain the expected value and see the return on investment. Progress on Salem is nearly 75% complete, while Hope Creek is 20% complete. To data, PSE and G has experienced success in the use of DBDs in areas such as development of plant modifications, development of the reliability-centered maintenance program, procedure upgrades, improved document retrieval, resolution of regulatory issues, and training. The paper examines the design basis development process, supporting projects, and expected improvements in plant operations as a result of these efforts

  1. Online operations optimization of waste incineration plants. Phase 3: Control concept and demonstration; Online driftsoptimering af affaldsfyrede anlaeg. Fase 3: Reguleringskoncept og demonstration. Hovedrapport ver. C

    Energy Technology Data Exchange (ETDEWEB)

    Boecher Poulsen, K.; Rassing Stoltze, K.; Solberg, B.; Hansen, Lars Henrik (DONG Energy (Denmark)); Cramer, J.; Andreasen, L.B. (FORCE Technology (Denmark)); Nymann Thomsen, S.; West, F. (Babcock and Wilcox Voelund (Denmark)); Clausen, S.; Fateev, A. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy, Roskilde (Denmark))

    2010-06-15

    have been considered, partly to assess the measuring campaign method and partly to find good MIMO models which can be used for more advanced control design, eg MPC. The dynamic characteristics form the basis of the control strategy design, and as valuable knowledge of the key parameters from two widely different plants is available, it is certain that the optimisation work is generic. A waste density soft sensor has been implemented and tested via radar measurement at Haderslev CHP Plant. The density soft sensor is used to adjust the pusher speed, thus reducing compressibility interruptions considerably. By doing so the control process does not have to wait for the modulus to increase or O{sub 2} and steam flow to be affected before a waste flow deviation can be detected. Based on measuring data it is obvious that the density control stabilises the waste flow to the furnace. A modulus control concept has been developed for coordination of several incineration zones, among other things by simulation. Three concepts have been selected, implemented and commissioned at Haderslev CHP Plant. One concept has been selected, and measuring data show that the modulus is considerably stabilised on activation. The concept combines the advantages with monitoring of the feed grate for overloading and at the same time maintenance of the incineration zone. However, in periods the concept is in conflict with steam flow and O{sub 2} which require manual operator intervention. Unfortunately, it was not possible to finalise and commission the overall control concept within the financial scope of the phase 3 project, because the maturation of the NIR cameras and the work with identification of the vast amount of data from the Reno-Nord tests exceeded the budget. Parts of the control concept have been commissioned and tested, but not the overall control concept as a whole. At the end of the project, it was decided to emphasise data processing and the theoretical mapping of the furnace

  2. Higs-instrument: design and demonstration of a high performance gas concentration imager

    Science.gov (United States)

    Verlaan, A. L.; Klop, W. A.; Visser, H.; van Brug, H.; Human, J.

    2017-09-01

    Climate change and environmental conditions are high on the political agenda of international governments. Laws and regulations are being setup all around the world to improve the air quality and to reduce the impact. The growth of a number of trace gasses, including CO2, Methane and NOx are especially interesting due to their environmental impact. The regulations made are being based on both models and measurements of the trend of those trace gases over the years. Now the regulations are in place also enforcement and therewith measurements become more and more important. Instruments enabling high spectral and spatial resolution as well as high accurate measurements of trace gases are required to deliver the necessary inputs. Nowadays those measurements are usually performed by space based spectrometers. The requirement for high spectral resolution and measurement accuracy significantly increases the size of the instruments. As a result the instrument and satellite becomes very expensive to develop and to launch. Specialized instruments with a small volume and the required performance will offer significant advantages in both cost and performance. Huib's Innovative Gas Sensor (HIGS, named after its inventor Huib Visser), currently being developed at TNO is an instrument that achieves exactly that. Designed to measure only a single gas concentration, opposed to deriving it from a spectrum, it achieves high performance within a small design volume. The instrument enables instantaneous imaging of the gas distribution of the selected gas. An instrument demonstrator has been developed for NO2 detection. Laboratory measurements proved the measurement technique to be successful. An on-sky measurement campaign is in preparation. This paper addresses both the instrument design as well as the demonstrated performances.

  3. Characterization and Design of Digital Pointing Subsystem for Optical Communication Demonstrator

    Science.gov (United States)

    Racho, C.; Portillo, A.

    1998-01-01

    The Optical Communications Demonstrator (OCD) is a laboratory-based lasercom demonstration terminal designed to validate several key technologies, including beacon acquisition, high bandwidth tracking, precision bearn pointing, and point-ahead compensation functions. It has been under active development over the past few years. The instrument uses a CCD array detector for both spatial acquisition and high-bandwidth tracking, and a fiber coupled laser transmitter. The array detector tracking concept provides wide field-of-view acquisition and permits effective platform jitter compensation and point-ahead control using only one steering mirror. This paper describes the detailed design and characterization of the digital control loop system which includes the Fast Steering Mirror (FSM), the CCD image tracker, and the associated electronics. The objective is to improve the overall system performance using laboratory measured data. The. design of the digital control loop is based on a linear time invariant open loop model. The closed loop performance is predicted using the theoretical model. With the digital filter programmed into the OCD control software, data is collected to verify the predictions. This paper presents the results of the, system modeling and performance analysis. It has been shown that measurement data closely matches theoretical predictions. An important part of the laser communication experiment is the ability of FSM to track the laser beacon within the. required tolerances. The pointing must be maintained to an accuracy that is much smaller than the transmit signal beamwidth. For an earth orbit distance, the system must be able to track the receiving station to within a few microradians. The failure. to do so will result in a severely degraded system performance.

  4. Design of nuclear power generation plants adopting model engineering method

    International Nuclear Information System (INIS)

    Waki, Masato

    1983-01-01

    The utilization of model engineering as the method of design has begun about ten years ago in nuclear power generation plants. By this method, the result of design can be confirmed three-dimensionally before actual production, and it is the quick and sure method to meet the various needs in design promptly. The adoption of models aims mainly at the improvement of the quality of design since the high safety is required for nuclear power plants in spite of the complex structure. The layout of nuclear power plants and piping design require the model engineering to arrange rationally enormous quantity of things in a limited period. As the method of model engineering, there are the use of check models and of design models, and recently, the latter method has been mainly taken. The procedure of manufacturing models and engineering is explained. After model engineering has been completed, the model information must be expressed in drawings, and the automation of this process has been attempted by various methods. The computer processing of design is in progress, and its role is explained (CAD system). (Kako, I.)

  5. Osiris and SOMBRERO inertial confinement fusion power plant designs

    International Nuclear Information System (INIS)

    Meier, W.R.; Bieri, R.L.; Monsler, M.J.

    1992-03-01

    The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs

  6. Nuclear power plant C and I design verification by simulation

    International Nuclear Information System (INIS)

    Storm, Joachim; Yu, Kim; Lee, D.Y

    2003-01-01

    An important part of the Advanced Boiling Water Reactor (ABWR) in the Taiwan NPP Lungmen Units no.1 and no.2 is the Full Scope Simulator (FSS). The simulator was to be built according to design data and therefore, apart from the training aspect, a major part of the development is to apply a simulation based test bed for the verification, validation and improvement of plant design in the control and instrumentation (C and I) areas of unit control room equipment, operator Man Machine Interface (MMI), process computer functions and plant procedures. Furthermore the Full Scope Simulator will be used after that to allow proper training of the plant operators two years before Unit no.1 fuel load. The article describes scope, methods and results of the advanced verification and validation process and highlights the advantages of test bed simulation for real power plant design and implementation. Subsequent application of advanced simulation software tools like instrumentation and control translators, graphical model builders, process models, graphical on-line test tools and screen based or projected soft panels, allowed a team to fulfil the task of C and I verification in time before the implementation of the Distributed Control and Information System (DCIS) started. An additional area of activity was the Human Factors Engineering (HFE) for the operator MMI. Due to the fact that the ABWR design incorporates a display-based operation with most of the plant components, a dedicated verification and validation process is required by NUREG-0711. In order to support this activity an engineering test system had been installed for all the necessary HFE investigations. All detected improvements had been properly documented and used to update the plant design documentation by a defined process. The Full Scope Simulator (FSS) with hard panels and stimulated digital control and information system are in the final acceptance test process with the end customer, Taiwan Power Company

  7. SusDesign - An approach for a sustainable process system design and its application to a thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Abul Hossain, K; Khan, F; Hawboldt, K [Mem University of Newfoundland, St John, NF (Canada). Faculty of Engineering & Applied Science

    2010-10-15

    This paper presents a structured process design approach, SusDesign, for the sustainable development of process systems. At each level of process design, design alternatives are generated using a number of thermodynamic tools and applying pollution prevention strategies followed by analysis, evaluation and screening processes for the selection of potential design options. The evaluation and optimization are carried out based on an integrated environmental and cost potential (IECP) index, which has been estimated with the IECP tool. The present paper also describes a flowsheet optimization technique developed using different thermodynamic tools such as exergy/energy analysis, heat and mass integration, and cogeneration/trigeneration in a systematic manner. The proposed SusDesign approach has been successfully implemented in designing a 30 MW thermal power plant. In the case study, the IECP tool has been set up in Aspen HYSYS process simulator to carry out the analysis, evaluation and screening of design alternatives. The application of this approach has developed an efficient, cost effective and environmentally friendly thermal system design with an overall thermal efficiency of 70% and CO{sub 2} and NO emissions of 0.28 kg/kW h and 0.2 g/kW h respectively. The cost of power generation is estimated as 4 cents kWh. These achievements are significant compared to the conventional thermal power plant, which demonstrates the potential of the SusDesign approach for the sustainable development of process systems.

  8. SusDesign - An approach for a sustainable process system design and its application to a thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hossain, Khandoker Abul [Faculty of Engineering and Applied Science Memorial University of Newfoundland, St. John' s, NL, A1B 3X5 (Canada); Khan, Faisal [Faculty of Engineering and Applied Science Memorial University of Newfoundland, St. John' s, NL, A1B 3X5 (Canada); Hawboldt, Kelly [Faculty of Engineering and Applied Science Memorial University of Newfoundland, St. John' s, NL, A1B 3X5 (Canada)

    2010-10-15

    This paper presents a structured process design approach, SusDesign, for the sustainable development of process systems. At each level of process design, design alternatives are generated using a number of thermodynamic tools and applying pollution prevention strategies followed by analysis, evaluation and screening processes for the selection of potential design options. The evaluation and optimization are carried out based on an integrated environmental and cost potential (IECP) index, which has been estimated with the IECP tool. The present paper also describes a flowsheet optimization technique developed using different thermodynamic tools such as exergy/energy analysis, heat and mass integration, and cogeneration/trigeneration in a systematic manner. The proposed SusDesign approach has been successfully implemented in designing a 30 MW thermal power plant. In the case study, the IECP tool has been set up in Aspen HYSYS process simulator to carry out the analysis, evaluation and screening of design alternatives. The application of this approach has developed an efficient, cost effective and environmentally friendly thermal system design with an overall thermal efficiency of 70% and CO{sub 2} and NO emissions of 0.28 kg/kW h and 0.2 g/kW h respectively. The cost of power generation is estimated as 4 cents /kW h. These achievements are significant compared to the conventional thermal power plant, which demonstrates the potential of the SusDesign approach for the sustainable development of process systems.

  9. SusDesign - An approach for a sustainable process system design and its application to a thermal power plant

    International Nuclear Information System (INIS)

    Hossain, Khandoker Abul; Khan, Faisal; Hawboldt, Kelly

    2010-01-01

    This paper presents a structured process design approach, SusDesign, for the sustainable development of process systems. At each level of process design, design alternatives are generated using a number of thermodynamic tools and applying pollution prevention strategies followed by analysis, evaluation and screening processes for the selection of potential design options. The evaluation and optimization are carried out based on an integrated environmental and cost potential (IECP) index, which has been estimated with the IECP tool. The present paper also describes a flowsheet optimization technique developed using different thermodynamic tools such as exergy/energy analysis, heat and mass integration, and cogeneration/trigeneration in a systematic manner. The proposed SusDesign approach has been successfully implemented in designing a 30 MW thermal power plant. In the case study, the IECP tool has been set up in Aspen HYSYS process simulator to carry out the analysis, evaluation and screening of design alternatives. The application of this approach has developed an efficient, cost effective and environmentally friendly thermal system design with an overall thermal efficiency of 70% and CO 2 and NO emissions of 0.28 kg/kW h and 0.2 g/kW h respectively. The cost of power generation is estimated as 4 cents /kW h. These achievements are significant compared to the conventional thermal power plant, which demonstrates the potential of the SusDesign approach for the sustainable development of process systems.

  10. Design and analysis of electrical energy storage demonstration projects on UK distribution networks

    International Nuclear Information System (INIS)

    Lyons, P.F.; Wade, N.S.; Jiang, T.; Taylor, P.C.; Hashiesh, F.; Michel, M.; Miller, D.

    2015-01-01

    Highlights: • Results of an EES system demonstration project carried out in the UK. • Approaches to the design of trials for EES and observation on their application. • A formalised methodology for analysis of smart grids trials. • Validated models of energy storage. • Capability of EES to connect larger quantities of heat pumps and PV is evaluated. - Abstract: The UK government’s CO 2 emissions targets will require electrification of much of the country’s infrastructure with low carbon technologies such as photovoltaic panels, electric vehicles and heat pumps. The large scale proliferation of these technologies will necessitate major changes to the planning and operation of distribution networks. Distribution network operators are trialling electrical energy storage (EES) across their networks to increase their understanding of the contribution that it can make to enable the expected paradigm shift in generation and consumption of electricity. In order to evaluate a range of applications for EES, including voltage control and power flow management, installations have taken place at various distribution network locations and voltage levels. This article reports on trial design approaches and their application to a UK trial of an EES system to ensure broad applicability of the results. Results from these trials of an EES system, low carbon technologies and trial distribution networks are used to develop validated power system models. These models are used to evaluate, using a formalised methodology, the impact that EES could have on the design and operation of future distribution networks

  11. Design of analog front-ends for the RD53 demonstrator chip

    CERN Document Server

    Gaioni, L; Nodari, B; Manghisoni, M; Re, V; Traversi, G; Barbero, M B; Fougeron, D; Gensolen, F; Godiot, S; Menouni, M; Pangaud, P; Rozanov, A; Wang, A; Bomben, M; Calderini, G; Crescioli, F; Le Dortz, O; Marchiori, G; Dzahini, D; Rarbi, F E; Gaglione, R; Gonella, L; Hemperek, T; Huegging, F; Karagounis, M; Kishishita, T; Krueger, H; Rymaszewski, P; Wermes, N; Ciciriello, F; Corsi, F; Marzocca, C; De Robertis, G; Loddo, F; Licciulli, F; Andreazza, A; Liberali, V; Shojaii, S; Stabile, A; Bagatin, M; Bisello, D; Mattiazzo, S; Ding, L; Gerardin, S; Giubilato, P; Neviani, A; Paccagnella, A; Vogrig, D; Wyss, J; Bacchetta, N; Della Casa, G; Demaria, N; Mazza, G; Rivetti, A; Da Rocha Rolo, M D; Comotti, D; Ratti, L; Vacchi, C; Beccherle, R; Bellazzini, R; Magazzu, G; Minuti, M; Morsani, F; Palla, F; Poulios, S; Fanucci, L; Rizzi, A; Saponara, S; Androsov, K; Bilei, G M; Menichelli, M; Conti, E; Marconi, S; Passeri, D; Placidi, P; Monteil, E; Pacher, L; Paternò, A; Gajanana, D; Gromov, V; Hessey, N; Kluit, R; Zivkovic, V; Havranek, M; Janoska, Z; Marcisovsky, M; Neue, G; Tomasek, L; Kafka, V; Sicho, P; Vrba, V; Vila, I; Lopez-Morillo, E; Aguirre, M A; Palomo, F R; Muñoz, F; Abbaneo, D; Christiansen, J; Dannheim, D; Dobos, D; Linssen, L; Pernegger, H; Valerio, P; Alipour Tehrani, N; Bell, S; Prydderch, M L; Thomas, S; Christian, D C; Fahim, F; Hoff, J; Lipton, R; Liu, T; Zimmerman, T; Garcia-Sciveres, M; Gnani, D; Mekkaoui, A; Gorelov, I; Hoeferkamp, M; Seidel, S; Toms, K; De Witt, J N; Grillo, A

    2017-01-01

    The RD53 collaboration is developing a large scale pixel front-end chip, which will be a tool to evaluate the performance of 65 nm CMOS technology in view of its application to the readout of the innermost detector layers of ATLAS and CMS at the HL-LHC. Experimental results of the characterization of small prototypes will be discussed in the frame of the design work that is currently leading to the development of the large scale demonstrator chip RD53A to be submitted in early 2017. The paper is focused on the analog processors developed in the framework of the RD53 collaboration, including three time over threshold front-ends, designed by INFN Torino and Pavia, University of Bergamo and LBNL and a zero dead time front-end based on flash ADC designed by a joint collaboration between the Fermilab and INFN. The paper will also discuss the radiation tolerance features of the front-end channels, which were exposed to up to 800 Mrad of total ionizing dose to reproduce the system operation in the actual experiment.

  12. Hanford Waste Vitrification Plant technical background document for toxics best available control technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    None

    1992-10-01

    This document provides information on toxic air pollutant emissions to support the Notice of Construction for the proposed Hanford Waste Vitrification Plant (HWVP) to be built at the the Department of Energy Hanford Site near Richland, Washington. Because approval must be received prior to initiating construction of the facility, state and federal Clean Air Act Notices of construction are being prepared along with necessary support documentation.

  13. Hanford Waste Vitrification Plant technical background document for toxics best available control technology demonstration

    International Nuclear Information System (INIS)

    1992-10-01

    This document provides information on toxic air pollutant emissions to support the Notice of Construction for the proposed Hanford Waste Vitrification Plant (HWVP) to be built at the the Department of Energy Hanford Site near Richland, Washington. Because approval must be received prior to initiating construction of the facility, state and federal Clean Air Act Notices of construction are being prepared along with necessary support documentation

  14. Great gas plants : these five natural gas processing facilities demonstrate decades of top-flight technology

    Energy Technology Data Exchange (ETDEWEB)

    Byfield, M.

    2010-07-15

    The natural gas purification and pipeline sector is a major economic driver in Canada. Gas processing facilities are growing in number, and several large gas projects are being planned for future construction in the western provinces. This article outlined 5 gas plants in order to illustrate the sector's history and breadth in Canada. The Shell Jumping Pound gas complex was constructed in 1951 after a sulfur-rich gas discovery near Calgary in 1944. The Empress Straddle plant was built in 1971 in southeastern Alberta and is one of the largest single industrial consumers of electrical power in the province. The Fort Nelson gas processing plant is North America's largest sour gas processing facility. The Shell Caroline complex was built 1993. The Sable offshore energy project is located on the coast of Nova Scotia to handle gas produced from the Thebaud wells. A consortium is now considering the development of new gas fields in the Sable area. 5 figs.

  15. Evaluation of masonry wall design at nuclear power plants

    International Nuclear Information System (INIS)

    Con, V.N.; Subramonian, N.; Chokshi, N.

    1983-01-01

    The structural integrity of safety-related masonry walls in operating nuclear power plants may not be maintained when subjected to certain loads and load combinations. The paper presents some findings based upon the review of the design and analysis procedures used by the licensees in the reevaluation of safety-related masonry walls. The design criteria developed by the Structural Engineering Branch (SEB) of the United States Nuclear Regulatory Commission (NRC) along with other standard codes such as the Uniform Building Code, ACI 531-79, ATC 3-06, and NCMA were used as guidance in evaluating the design criteria developed by the licensees. The paper deals with the following subject areas: loads and load combinations, allowable stresses, analytical procedures, and modification methods. The paper concludes that, in general, the masonry walls in nuclear power plants comply with the working stress design requirements. In some cases, certain nonlinear analysis methods were used. The applicability of these methods is discussed. (orig.)

  16. Accounting for maintenance in the design of nuclear power plants

    International Nuclear Information System (INIS)

    Meuwisse, C.; Martin-Mattei, C.; Hamon, L.

    1997-01-01

    The objective of the CIDEM project (French acronym for Design Integrating Availability, Operating Experience and Maintenance) is to control the per-kW production cost of future Electricite de France REP 2000 nuclear plants. In particular, such cost control requires accounting for maintenance and logistic support from the time of design of the future installations. This technical and economic optimization is based on assessment and comparison of possible choices in terms of materials and maintenance, and on the search for potential improvements. In the Basic Design phase, the method involves identifying reference components which are supposed highly similar to the components to be designed. In the Construction phase, it culminates in the early definition of a detailed maintenance and support plant. To be effective, the approach requires realistic tailoring of the studies to be undertaken, and the participation of multidisciplinary teams working in the framework of concurrent engineering. (author)

  17. Design aid system for nuclear power plant instrumentations

    International Nuclear Information System (INIS)

    Hattori, Yoshiaki; Ito, Toshiichiro; Fujii, Makoto; Shimada, Nobuhide.

    1987-01-01

    Purpose: To enable to provide design aid for the nuclear power plant instrumentation of high reliability with the minimum cost while eliminating unrequired condition even if there are no data for the ground of the instrumentation design. Constitution: The information data base for the design of process radiation ray monitors are administrated by a data base administration device. The conditions to be satisfied in the process radiation monitors designed based on the data for the circumstances where particular predetermined process radiation monitors are installed, are derived by deduction using information obtained from the data base by way of the data base administration device. The derived design conditions are displayed and the optimum conditions are again reduced and displayed. In this way, the designers are assisted such that optimum designs can be obtained while sufficiently satisfying the safety and also in view of the cost. (Kamimura, M.)

  18. Designing for nuclear power plant maintainability and operability

    International Nuclear Information System (INIS)

    Pedersen, T.J.

    1998-01-01

    Experience has shown that maintenance and operability aspects must be addressed in the design work. ABB Atom has since long an ambition of achieving optimised, overall plant designs, and efficient feedback of growing operating experience has stepwise eliminated shortcomings, and yielded better and better plant operating performances. The records of the plants of the latest design versions are very good; four units in Sweden have operated at an energy availability of 90.1%, and the two Olkiluoto units in Finland at a load factor of 92.7%, over the last decade. The occupational radiation exposures have also been at a low level. The possibilities for implementing 'lessons learned' in existing plants are obviously limited by practical constraints. In Finland and Sweden, significant modernisations are still underway, however, involving replacement of mechanical equipment, and upgrading and backfitting of I and C systems on a large scale, in most of the plants. The BWR 90 design focuses on meeting requirements from utilities as well as new regulatory requirements, with a particular emphasis on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimisation of buildings and containment to decrease construction time and costs, and selection of materials as well as maintenance of operating procedures to reduce radiation exposures even further. The BWR 90 design was offered to Finland in the early 1990s, but development work continues. It has been selected by a number of European utilities for assessing its conformance with the European Utility Requirements (EUR), aiming at a specific EUR Volume 3 for the BWR 90. Some characteristics of the ABB BWRs, with emphasis on features of importance for achieving improved economy and enhanced safety, are described below. (author)

  19. Priorities in the design of chemical shops at coke plants

    Energy Technology Data Exchange (ETDEWEB)

    V.I. Rudyka; Y.E. Zingerman; V.V. Grabko; L.A. Kazak [Giprokoks, the State Institute for the Design of Coke-Industry Enterprises, Kharkov (Ukraine)

    2009-07-15

    Recent trends in the design of chemical equipment at coke plants are described, through the lens of experience at Giprokoks. The main priorities were to improve the removal of impurities from coke oven gas; to improve equipment design on the basis of new materials; to reduce reagent consumption; to reduce the materials and energy consumed in the construction of new equipment; and to minimize impacts on the environment and worker health. Some technological equipment is briefly characterized.

  20. Architect engineer balance-of-plant radiological design considerations

    International Nuclear Information System (INIS)

    Piccot, A.R.

    1975-01-01

    Methods which are or may be used by Architect Engineers in dealing with the problems of radiological safety in the design of a nuclear power plant are discussed. The bases and basic requirements for a radiation protection program are briefly noted. Requirements in the areas of planning, organization, responsibilities and implementation of radiation protection are discussed. Lists of safety tasks which should be performed during the various design phases are presented

  1. Demonstration of Active Power Controls by Utility-Scale PV Power Plant in an Island Grid: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Gevorgian, Vahan; O' Neill, Barbara

    2017-02-01

    The National Renewable Energy Laboratory (NREL), AES, and the Puerto Rico Electric Power Authority conducted a demonstration project on a utility-scale photovoltaic (PV) plant to test the viability of providing important ancillary services from this facility. As solar generation increases globally, there is a need for innovation and increased operational flexibility. A typical PV power plant consists of multiple power electronic inverters and can contribute to grid stability and reliability through sophisticated 'grid-friendly' controls. In this way, it may mitigate the impact of its variability on the grid and contribute to important system requirements more like traditional generators. In 2015, testing was completed on a 20-MW AES plant in Puerto Rico, and a large amount of test data was produced and analyzed that demonstrates the ability of PV power plants to provide various types of new grid-friendly controls. This data showed how active power controls can leverage PV's value from being simply an intermittent energy resource to providing additional ancillary services for an isolated island grid. Specifically, the tests conducted included PV plant participation in automatic generation control, provision of droop response, and fast frequency response.

  2. Kentucky Department for Natural Resources and Environmental Protection permit application for air contaminant source: SRC-I demonstration plant, Newman, Kentucky. Appendix D. Impact assessment. [Demonstration plant at Newman, KY

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-11-21

    In appendix D, the air quality condition for various pollutants in the areas surrounding the proposed demonstration plant site is given with respect to attainment or non-attainment of US EPA regulations. The minimum pollutant emission rates for these regulated and for several other pollutants are given. Then the estimated emission rates from the proposed plant are given for a dozen pollutants which exceed these limits and therefore require an ambient air quality analysis. This involves taking into account the estimated emission of these pollutants from the proposed plant and from other sources in the surrounding area. Finally, background data from the surrounding area including meteorological data and sampling of regulated pollutants are given. (LTN)

  3. Safety demonstration tests of postulated solvent fire accidents in extraction process of a fuel reprocessing plant, (2)

    International Nuclear Information System (INIS)

    Tukamoto, Michio; Takada, Junichi; Koike, Tadao; Nishio, Gunji; Uno, Seiichiro; Kamoshida, Atsusi; Watanabe, Hironori; Hashimoto, Kazuichiro; Kitani, Susumu.

    1992-03-01

    Demonstration tests of hypothetical solvent fire in an extraction process of the reprocessing plant were carried out from 1984 to 1985 in JAERI, focusing on the confinement of radioactive materials during the fire by a large-scale fire facility (FFF) to evaluate the safety of air-ventilation system in the plant. Fire data from the demonstration test were obtained by focusing on fire behavior at cells and ducts in the ventilation system, smoke generation during the fire, transport and deposition of smoke containing simulated radioactive species in the ventilation system, confinement of radioactive materials, and integrity of HEPA filters by using the FFF simulating an air-ventilation system of the reference reprocessing plant in Japan. The present report is published in a series of the report Phase I (JAERI-M 91-145) of the demonstration test. Test results in the report will be used for the verification of a computer code FACE to evaluate the safety of postulated fire accidents in the reprocessing plant. (author)

  4. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. This document, Volume II, contains Appendices 2, 3, 5, 6, and 7 covering the following topics: Eastern North American Empirical Ground Motion Data; Examination of Variance of Seismographic Network Data; Soil Amplification and Vertical-to-Horizontal Ratios from Analysis of Strong Motion Data From Active Tectonic Regions; Revision and Calibration of Ou and Herrmann Method; Generalized Ray Procedure for Modeling Ground Motion Attenuation; Crustal Models for Velocity Regionalization; Depth Distribution Models; Development of Generic Site Effects Model; Validation and Comparison of One-Dimensional Site Response Methodologies; Plots of Amplification Factors; Assessment of Coupling Between Vertical ampersand Horizontal Motions in Nonlinear Site Response Analysis; and Modeling of Dynamic Soil Properties

  5. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions. Volume 2, Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-18

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake`s ground motion is a function of the earthquake`s magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. This document, Volume II, contains Appendices 2, 3, 5, 6, and 7 covering the following topics: Eastern North American Empirical Ground Motion Data; Examination of Variance of Seismographic Network Data; Soil Amplification and Vertical-to-Horizontal Ratios from Analysis of Strong Motion Data From Active Tectonic Regions; Revision and Calibration of Ou and Herrmann Method; Generalized Ray Procedure for Modeling Ground Motion Attenuation; Crustal Models for Velocity Regionalization; Depth Distribution Models; Development of Generic Site Effects Model; Validation and Comparison of One-Dimensional Site Response Methodologies; Plots of Amplification Factors; Assessment of Coupling Between Vertical & Horizontal Motions in Nonlinear Site Response Analysis; and Modeling of Dynamic Soil Properties.

  6. New Clean Air Act complicates power plant operation, design

    International Nuclear Information System (INIS)

    Smock, R.W.

    1991-01-01

    In November the president signed into law the new Clean Air Act, ushering in a new era in the power generation industry. This paper reviews the six important sections of the Clean Air Act and their impact on power plant operation and design

  7. Some engineering considerations when designing centrifuge enrichment plants

    International Nuclear Information System (INIS)

    Edwards, T.T.

    1982-01-01

    A review is given of the three main areas where flexibility is needed in the design of centrifuge enrichment plants. These are: the need to cope with market requirements, the limitations imposed by currently available centrifuges and ever advancing centrifuge technology. Details of BNFL's experience with centrifuge enrichment at Capenhurst are presented. (U.K.)

  8. The Design and Manufacturing of Essential oil Distillation Plant for ...

    African Journals Online (AJOL)

    Choice-Academy

    industry in the country do not have the capacity to manufacture the complete distillation plant system with the required precision for standard quality of oil at affordable cost. Thus, the design and the experiment in the use of a prototype small size distillation unit showed that the technology is appropriate for essential oil ...

  9. Preconceptual design study for solidifying high-level waste: West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Hill, O.F.

    1981-04-01

    This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass

  10. Suitability of second pass RO as a substitute for high quality MSF product water in Nuclear Desalination Demonstration Plant

    International Nuclear Information System (INIS)

    Murugan, V.; Venkatesh, P.; Balasubramanian, C.; Nagaraj, R.; Yadav, Manoj Kumar; Prabhakar, S.; Tewari, P.K.

    2012-01-01

    Nuclear Desalination Demonstration Plant at Kalpakkam consists of both Multi Stage Flash Distillation (MSF) and Seawater Reverse Osmosis (SWRO) process to produce desalinated water. It supplies part of highly pure water from MSF to Madras Atomic Power Station for its boiler feed requirements and remaining water is blend with SWRO product water and sent to other common facilities located inside Kalpakkam campus. A critical techno-economic analysis is carried out to find out the suitability of second pass RO to sustain the availability of highly pure water in case of MSF plant shutdown. (author)

  11. Design and demonstration of adiabatic quantum-flux-parametron logic circuits with superconductor magnetic shields

    International Nuclear Information System (INIS)

    Inoue, Kenta; Narama, Tatsuya; Yamanashi, Yuki; Yoshikawa, Nobuyuki; Takeuchi, Naoki

    2015-01-01

    Adiabatic quantum-flux-parametron (AQFP) logic is an energy-efficient superconductor logic with zero static power and very small dynamic power due to adiabatic switching operations. In order to build large-scale digital circuits, we built AQFP logic cells using superconductor magnetic shields, which are necessary in order to avoid unwanted magnetic couplings between the cells and excitation currents. In preliminary experimental tests, we confirmed that the unwanted coupling became negligibly small thanks to the superconductor shields. As a demonstration, we designed a four-to-one multiplexor and a 16-junction full adder using the shielded logic cells. In both circuits, we confirmed correct logic operations with wide operation margins of excitation currents. These results indicate that large-scale AQFP digital circuits can be realized using the shielded logic cells. (paper)

  12. Wind tunnel testing of the DeepWind demonstrator in design and tilted operating conditions

    DEFF Research Database (Denmark)

    Battistia, L.; Benini, E.; Brighenti, A.

    2016-01-01

    The DeepWind Project aims at investigating the feasibility of a new floating vertical-axis wind turbine (VAWT) concept, whose purpose is to exploit wind resources at deep-water offshore sites. The results of an extensive experimental campaign on the DeepWind reduced scale demonstrator are here...... was installed on a high precision test bench, whose axis was suitable to be inclined up to 15° with respect to the design (i.e. upright) operating condition. The experiments were performed at the large scale, high speed wind tunnel of the Politecnico di Milano (Italy), using a “free jet” (open channel...... presented for different wind speeds and rotor angular velocities, including also skewed flow operation due to a tilted rotor arrangement. To accomplish this, after being instrumented to measure aerodynamic power and thrust (both in streamwise and transversal directions), a troposkien three-bladed rotor...

  13. A pilot plant demonstration of the vitrification of radioactive solutions using microwave power

    International Nuclear Information System (INIS)

    Morrell, M.S.; Hardwick, W.H.; Murphy, V.; Wace, P.F.

    1986-01-01

    A process has been developed that exploits the characteristics of microwave heating for the vitrification of high-level radioactive liquid waste. This process, microwave vitrification, has been successfully operated at pilot plant scale in an active cell using simulated liquid waste containing several curies of radioactivity. Excellent decontamination factors have been achieved for both volatiles and nonvolatiles with an average ruthenium decontamination factor of 490 and a gross alpha emitter decontamination factor of 100,000. Almost all the radioactivity is incorporated in a glass block

  14. In-can melting demonstration of wastes from the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Bjorklund, W.J.; Chick, L.A.; Hollis, H.H.; Mellinger, G.B.; Nelson, T.A.; Petkus, L.L.

    1980-07-01

    The immobilization of Idaho Chemical Processing Plant (ICPP) zirconia calcine using Idaho glass composition (ICPP-127) was evaluated at Pacific Northwest Laboratory (PNL) in two engineering-scale in-can melter tests. The glass was initially characterized in the laboratory to verify processing parameters. Glass was then produced in a pilot-scale melter and then in a full-scale melter to evaluate the processing and the resultant product. Potential corrosion problems were identified with the glass and some processing problems were encountered, but neither is insurmountable. The product is a durable leach-resistant glass. The glass appears to be nonhomogeneous, but chemically it is quite uniform

  15. Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    Bambang Galung Susanto

    2007-01-01

    The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO 2 /year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

  16. Design and Demonstration of an Advanced Agricultural Feedstock Supply System for Lignocellulosic Bioenergy Production

    Energy Technology Data Exchange (ETDEWEB)

    Clark, Timothy C. [Antares Group Inc., Lanham, MD (United States); Comer, Kevin S. [Antares Group Inc., Lanham, MD (United States); Belden, Jr., William S. [Antares Group Inc., Lanham, MD (United States)

    2016-04-30

    This three-year project developed and demonstrated four innovative, first-of-their-kind pieces of equipment that are aimed at significantly reducing the cost of delivered herbaceous biomass. This equipment included a Self-Propelled Baler (SPB), a Bale Picking Truck (BPT), a Self-Loading Trailer (SLT), and a Heavy Crop Header for harvesting high yielding energy crops. This equipment was designed and fabricated during the first two years of the project and demonstrated on available crops (corn stover, wheat straw, and warm season grasses) across the nation, as available. Operational performance and cost data was collected and analyzed throughout the project to measure the costs of baseline harvesting (using conventional harvesting equipment) and advanced harvesting with the newly developed equipment. This data revealed that the project met its original goal of developing equipment that is realistically capable of reducing the cost of delivered biomass by $13 per dry ton. Each machine was tested after fabrication and put to the test in one or more commercial harvesting seasons. During these tests, operational flaws were found and fixed through upgrades and improvements. The first new SPB, BPT, and two new SLTs were ready for use during the 2013 harvest season. Since then, over 40 SLTs have been ordered and are currently under fabrication. All of the equipment will be commercially available to the industry as demand increases.

  17. Microgrid Design, Development and Demonstration - Final Report for Phase I and Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Bose, Sumit [GE Global Research Center, Niskayuna, NY (United States); Krok, Michael [GE Global Research Center, Niskayuna, NY (United States)

    2011-02-08

    This document constitutes GE’s final report for the Microgrid Design, Development and Demonstration program for DOE’s Office of Electricity Delivery and Energy Reliability, Award DE-FC02-05CH11349. It contains the final report for Phase I in Appendix I, and the results the work performed in Phase II. The program goal was to develop and demonstrate a Microgrid Energy Management (MEM) framework for a broad set of Microgrid applications that provides unified controls, protection, and energy management. This project contributed to the achievement of the U.S. Department of Energy’s Renewable and Distributed Systems Integration Program goals by developing a fully automated power delivery microgrid network that: - Reduces carbon emissions and emissions of other air pollutants through increased use of optimally dispatched renewable energy, - Increases asset use through integration of distributed systems, - Enhances reliability, security, and resiliency from microgrid applications in critical infrastructure protection, constrained areas of the electric grid, etc. - Improves system efficiency with on-site, distributed generation and improved economic efficiency through demand-side management.

  18. Curved sensors for compact high-resolution wide-field designs: prototype demonstration and optical characterization

    Science.gov (United States)

    Chambion, Bertrand; Gaschet, Christophe; Behaghel, Thibault; Vandeneynde, Aurélie; Caplet, Stéphane; Gétin, Stéphane; Henry, David; Hugot, Emmanuel; Jahn, Wilfried; Lombardo, Simona; Ferrari, Marc

    2018-02-01

    Over the recent years, a huge interest has grown for curved electronics, particularly for opto-electronics systems. Curved sensors help the correction of off-axis aberrations, such as Petzval Field Curvature, astigmatism, and bring significant optical and size benefits for imaging systems. In this paper, we first describe advantages of curved sensor and associated packaging process applied on a 1/1.8'' format 1.3Mpx global shutter CMOS sensor (Teledyne EV76C560) into its standard ceramic package with a spherical radius of curvature Rc=65mm and 55mm. The mechanical limits of the die are discussed (Finite Element Modelling and experimental), and electro-optical performances are investigated. Then, based on the monocentric optical architecture, we proposed a new design, compact and with a high resolution, developed specifically for a curved image sensor including optical optimization, tolerances, assembly and optical tests. Finally, a functional prototype is presented through a benchmark approach and compared to an existing standard optical system with same performances and a x2.5 reduction of length. The finality of this work was a functional prototype demonstration on the CEA-LETI during Photonics West 2018 conference. All these experiments and optical results demonstrate the feasibility and high performances of systems with curved sensors.

  19. Dry spent-fuel consolidation demonstration at the Barnwell Nuclear Fuel Plant (BNFP)

    International Nuclear Information System (INIS)

    Townes, G.A.

    1982-08-01

    Equipment for disassembling and canning (or encapsulating) spent fuel to allow more efficient storage is being developed and demonstrated at the BNFP. The program is aimed at dry disassembly of fuel to allow storage and shipment of fuel pins rather than full fuel assemblies. Results indicate that doubling the existing storage capacity or tripling the carrying capacity of existing transportation equipment is achievable. Disassembly has been demonstrated in the BNFP hot cells at rates of approx. 10 to 12 assemblies per day. 3 figures

  20. Preliminary design of a gas-cooled accelerator driven system demonstrator

    International Nuclear Information System (INIS)

    Giraud, B.; Poitevin, Y.; Ritter, G.

    2001-01-01

    At the present time, nuclear power appears to be the best solution for producing a large amount of electricity from both economical and ecological viewpoints, provided that acceptable answers to the nuclear waste concern are found. In France, this is the subject of the law 91-1381 (December 30 th , 1991). The transmutation of most of the long-lived radioactive wastes is a promising solution which could play a substantial role for the safety of the fuel cycle. Sub-critical Accelerator Driven System (ADS), coupling an ion accelerator and a sub-critical reactor, seems to have a high capacity for the fission of minor actinides and transmutation of long life fission products. The practicality on an industrial scale of partitioning and transmutation through ADS for reducing the amount of long life radio-nuclides has to be evaluated. It was recognised that the most efficient way, in terms of cost and planning, to conclusively assess the potential and the feasibility of a full scale industrial programme on ADS was to design and operate an ADS Demonstrator. The main ADS DF characteristics, defined within a joint working group, and reactor design features are described and justified. Then, main issues which call for research and development support are identified. (authors)

  1. Performance demonstration of a high-power space-reactor heat-pipe design

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Martinez, E.H.; Keddy, E.S.; Runyan, J.; Kemme, J.E.

    1983-01-01

    Performance of a 15.9-mm diam, 2-m long, artery heat pipe has been demonstrated at power levels to 22.6 kW and temperatures to 1500 0 K. The heat pipe employed lithium as a working fluid with distribution wicks and arteries fabricated from 400 mesh Mo-41 wt % Re screen. Molybdenum alloy (TZM) was used for the container. Peak axial power density attained in the testing was 19 kW/cm 2 at 1465 0 K. The corresponding radial flux density in the evaporator region of the heat pipe was 150 W/cm 2 . The extrapolated limit for the heat pipe at its 1500 0 K design point is 30 kW, corresponding to an axial flux density of 25 kW/cm 2 . Sonic and capillary limits for the design were investigated in the 1100 to 1500 0 K temperature range. Excellent agreement of measured and predicted temperature and power levels was observed

  2. Demonstration of improved vehicle fuel efficiency through innovative tire design, materials, and weight reduction technologies

    Energy Technology Data Exchange (ETDEWEB)

    Donley, Tim [Cooper Tire & Rubber Company Incorporated, Findlay, OH (United States)

    2014-12-31

    Cooper completed an investigation into new tire technology using a novel approach to develop and demonstrate a new class of fuel efficient tires using innovative materials technology and tire design concepts. The objective of this work was to develop a new class of fuel efficient tires, focused on the “replacement market” that would improve overall passenger vehicle fuel efficiency by 3% while lowering the overall tire weight by 20%. A further goal of this project was to accomplish the objectives while maintaining the traction and wear performance of the control tire. This program was designed to build on what has already been accomplished in the tire industry for rolling resistance based on the knowledge and general principles developed over the past decades. Cooper’s CS4 (Figure #1) premium broadline tire was chosen as the control tire for this program. For Cooper to achieve the goals of this project, the development of multiple technologies was necessary. Six technologies were chosen that are not currently being used in the tire industry at any significant level, but that showed excellent prospects in preliminary research. This development was divided into two phases. Phase I investigated six different technologies as individual components. Phase II then took a holistic approach by combining all the technologies that showed positive results during phase one development.

  3. Application of NASA Kennedy Space Center system assurance analysis methodology to nuclear power plant systems designs

    International Nuclear Information System (INIS)

    Page, D.W.

    1985-01-01

    The Kennedy Space Center (KSC) entered into an agreement with the Nuclear Regulatory Commission (NRC) to conduct a study to demonstrate the feasibility and practicality of applying the KSC System Assurance Analysis (SAA) methodology to nuclear power plant systems designs. In joint meetings of KSC and Duke Power personnel, an agreement was made to select to CATAWBA systems, the Containment Spray System and the Residual Heat Removal System, for the analyses. Duke Power provided KSC with a full set a Final Safety Analysis Reports as well as schematics for the two systems. During Phase I of the study the reliability analyses of the SAA were performed. During Phase II the hazard analyses were performed. The final product of Phase II is a handbook for implementing the SAA methodology into nuclear power plant systems designs. The purpose of this paper is to describe the SAA methodology as it applies to nuclear power plant systems designs and to discuss the feasibility of its application. The conclusion is drawn that nuclear power plant systems and aerospace ground support systems are similar in complexity and design and share common safety and reliability goals. The SAA methodology is readily adaptable to nuclear power plant designs because of it's practical application of existing and well known safety and reliability analytical techniques tied to an effective management information system

  4. Demonstration of Coupled Multievent Scenario at a Subject Nuclear Power Plant

    International Nuclear Information System (INIS)

    Coleman, Justin Leigh; Prescott, Steven Ralph; Smith, Curtis; Sampath, Ram

    2015-01-01

    This report discusses how to perform a coupled, seismic and flooding, multievent risk-informed analysis. Presented in the following sections are the need for multievent risk-informed analysis, the tools needed to perform the analysis, and an example of solving a demonstration problem.

  5. Demonstration using field collections that Argentina fall armyworm populations exhibit strain-specific host plant preferences

    Science.gov (United States)

    Spodoptera frugiperda, the fall armyworm, is a major economic pest throughout the Western Hemisphere of corn (maize), cotton, sorghum, and a variety of agricultural grasses and vegetable crops. Studies in the United States, the Caribbean, and Brazil demonstrated the existence of two subpopulations ...

  6. Short-cut design of small hydroelectric plants

    Energy Technology Data Exchange (ETDEWEB)

    Voros, N.G.; Kiranoudis, C.T.; Maroulis, Z.B. [National Technical Univ. of Athens, Dept. of Chemical Engineering, Athens (Greece)

    2000-04-01

    The problem of designing small hydroelectric plants has been properly analysed and addressed in terms of maximising the economic benefits of the investment. An appropriate empirical model describing hydroturbine efficiency was developed. An overall plant model was introduced by taking into account their construction characteristics and operational performance. The hydrogeographical characteristics for a wide range of sites have been appropriately analysed and a model that involves significant physical parameters has been developed. The design problem was formulated as a mathematical programming problem, and solved using appropriate programming techniques. The optimisation covered a wide range of site characteristics and three types of commercially available hydroturbines. The methodology introduced an empirical short-cut design equation for the determination of the optimum nominal flowrate of the hydroturbines and the estimation of the expected unit cost of electricity produced, as well as of the potential amount of annually recovered energy. (Author)

  7. Anthropometric data base for power plant design. Special report

    International Nuclear Information System (INIS)

    Parris, H.L.

    1981-07-01

    The primary study objective is to develop anthropometric data based upon the men and women who operate and maintain nuclear power plants. Age, stature, and weight information were obtained by a questionnaire survey of current operator and maintenance personnel, and the data extracted from the questionnaires were analyzed to derive body-size information for a number of anthropometric variables of interest to designers. Body-size information was developed separately for both men and women. Results achieved for the male population can be utilized by designers with a high level of confidence for the design of general workplaces. While the number of women respondents in the sample proved to be too small to derive results to which a similarly high level of reliability could be attached, the data can nevertheless be used as reasonable indicators of the probable body-size variability to be found among female power plant employees

  8. General design safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide provides the safety principles and the approach that have been used to implement the Code in the Safety Guides. These safety principles and the approach are tied closely to the safety analyses needed to assist the design process, and are used to verify the adequacy of nuclear power plant designs. This Guide also provides a framework for the use of other design Safety Guides. However, although it explains the principles on which the other Safety Guides are based, the requirements for specific applications of these principles are mostly found in the other Guides

  9. 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

  10. Manpower simulation for the power plant design engineering

    International Nuclear Information System (INIS)

    Moon, B.S.; Juhn, P.E.

    1982-01-01

    Some observation from the examination of actual manhour curves for the power design engineering obtained from Sargent and Lundy Engineers and of a few of the model curves proposed by Bechtel, are analyzed in this paper. A model curve representing typical design engineering manhour has been determined as probability density function for the Gamma Distribution. By means of this model curve, we strategically forecast the future engineering manpower requirements to meet the Covernment's long range nuclear power plan. As a sensitivity analysis, the directions for the localization of nuclear power plant design engineering, are studied in terms of the performance factor for the experienced versus inexperienced engineers. (Author)

  11. Construction and start-up of a 250 kW natural gas fueled MCFC demonstration power plant

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa, R.A.; Carter, J.; Rivera, R.; Otahal, J. [San Diego Gas & Electric, CA (United States)] [and others

    1996-12-31

    San Diego Gas & Electric (SDG&E) is participating with M-C Power in the development and commercialization program of their internally manifolded heat exchanger (IMHEX{reg_sign}) carbonate fuel cell technology. Development of the IMHEX technology base on the UNOCAL test facility resulted in the demonstration of a 250 kW thermally integrated power plant located at the Naval Air Station at Miramar, California. The members of the commercialization team lead by M-C Power (MCP) include Bechtel Corporation, Stewart & Stevenson Services, Inc., and Ishikawajima-Harima Heavy Industries (IHI). MCP produced the fuel cell stack, Bechtel was responsible for the process engineering including the control system, Stewart & Stevenson was responsible for packaging the process equipment in a skid (pumps, desulfurizer, gas heater, turbo, heat exchanger and stem generator), IHI produced a compact flat plate catalytic reformer operating on natural gas, and SDG&E assumed responsibility for plant construction, start-up and operation of the plant.

  12. The End of the Line, Preparing the Main Plant Process Building for Demolition at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Rowell, L.E.; Kurasch, D.H.; Hackett, M.; Gorsuch, G.; Sullivan, D.W.

    2009-01-01

    The West Valley Demonstration Project (WVDP) Act of 1980 authorized the Department of Energy to conduct a high-level radioactive waste management demonstration project at the site of the former Spent Fuel Reprocessing Plant in West Valley, New York to demonstrate solidification techniques to prepare high-level liquid waste for disposal. The reprocessing facility at this site was the only commercial NRC-licensed spent fuel reprocessing plant to have operated in the United States. The spent fuel reprocessing operations ended in 1972 and DoE's cleanup operations have been underway since 1982. High-level waste solidification was safely concluded in 2002 and follow-on activities at the site have been concentrated on facility decontamination and waste management and off-site disposal. Among the features that remain at the WVDP site is the highly-contaminated Main Plant Process Building (MPPB). The five-story reinforced concrete structure, which was formerly used to reprocess irradiated nuclear fuel, contains residual levels of contamination in some areas that prohibit safe human entry. DoE's long-range plans for the site include demolition of the MPPB. Current site contractor, West Valley Environmental Services LLC (WVES), while actively working to dismantle equipment and decontaminate areas inside the MPPB, has developed a conceptual two-phase plan for demolishing the structure that provides a cost-effective, lower-dose alternative to conventional demolition techniques. This paper discusses the current condition of the MPPB and the demolition-ready preparations conducted in the facility thus far. This paper also introduces the concept of a two-part surgical demolition plan that has been proposed and is being evaluated as a safe method of demolishing the structure. The practical applications that support feasibility for the demolition approach are being demonstrated through current work applications in the MPPB. The Inside-Out Demolition proposal for the MPPB is a safe

  13. Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

    Science.gov (United States)

    Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

    2015-04-01

    Preparedness of nuclear power plants to beyond design base external effects became high importance after 11th of March 2011 Great Tohoku Earthquakes. In case of some nuclear power plants constructed at the soft soil sites, liquefaction should be considered as a beyond design basis hazard. The consequences of liquefaction have to be analysed with the aim of definition of post-event plant condition, identification of plant vulnerabilities and planning the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The case of Nuclear Power Plant at Paks, Hungary is used as an example for demonstration of practical importance of the presented results and considerations. Contrary to the design, conservatism of the methodology for the evaluation of beyond design basis liquefaction effects for an operating plant has to be limited to a reasonable level. Consequently, applicability of all existing methods has to be considered for the best estimation. The adequacy and conclusiveness of the results is mainly limited by the epistemic uncertainty of the methods used for liquefaction hazard definition and definition of engineering parameters characterizing the consequences of liquefaction. The methods have to comply with controversial requirements. They have to be consistent and widely accepted and used in the practice. They have to be based on the comprehensive database. They have to provide basis for the evaluation of dominating engineering parameters that control the post-liquefaction response of the plant structures. Experience of Kashiwazaki-Kariwa plant hit by Niigata-ken Chuetsu-oki earthquake of 16 July 2007 and analysis of site conditions and plant layout at Paks plant have shown that the differential settlement is found to be the dominating effect in case considered. They have to be based on the probabilistic seismic hazard assessment and allow the integration into logic

  14. Genetic algorithms and Monte Carlo simulation for optimal plant design

    International Nuclear Information System (INIS)

    Cantoni, M.; Marseguerra, M.; Zio, E.

    2000-01-01

    We present an approach to the optimal plant design (choice of system layout and components) under conflicting safety and economic constraints, based upon the coupling of a Monte Carlo evaluation of plant operation with a Genetic Algorithms-maximization procedure. The Monte Carlo simulation model provides a flexible tool, which enables one to describe relevant aspects of plant design and operation, such as standby modes and deteriorating repairs, not easily captured by analytical models. The effects of deteriorating repairs are described by means of a modified Brown-Proschan model of imperfect repair which accounts for the possibility of an increased proneness to failure of a component after a repair. The transitions of a component from standby to active, and vice versa, are simulated using a multiplicative correlation model. The genetic algorithms procedure is demanded to optimize a profit function which accounts for the plant safety and economic performance and which is evaluated, for each possible design, by the above Monte Carlo simulation. In order to avoid an overwhelming use of computer time, for each potential solution proposed by the genetic algorithm, we perform only few hundreds Monte Carlo histories and, then, exploit the fact that during the genetic algorithm population evolution, the fit chromosomes appear repeatedly many times, so that the results for the solutions of interest (i.e. the best ones) attain statistical significance

  15. Probability sampling design in ethnobotanical surveys of medicinal plants

    Directory of Open Access Journals (Sweden)

    Mariano Martinez Espinosa

    2012-07-01

    Full Text Available Non-probability sampling design can be used in ethnobotanical surveys of medicinal plants. However, this method does not allow statistical inferences to be made from the data generated. The aim of this paper is to present a probability sampling design that is applicable in ethnobotanical studies of medicinal plants. The sampling design employed in the research titled "Ethnobotanical knowledge of medicinal plants used by traditional communities of Nossa Senhora Aparecida do Chumbo district (NSACD, Poconé, Mato Grosso, Brazil" was used as a case study. Probability sampling methods (simple random and stratified sampling were used in this study. In order to determine the sample size, the following data were considered: population size (N of 1179 families; confidence coefficient, 95%; sample error (d, 0.05; and a proportion (p, 0.5. The application of this sampling method resulted in a sample size (n of at least 290 families in the district. The present study concludes that probability sampling methods necessarily have to be employed in ethnobotanical studies of medicinal plants, particularly where statistical inferences have to be made using data obtained. This can be achieved by applying different existing probability sampling methods, or better still, a combination of such methods.

  16. Haida Gwaii / Queen Charlotte Islands demonstration tidal power plant feasibility study : summary results

    Energy Technology Data Exchange (ETDEWEB)

    Tu, A. [BC Hydro, Burnaby, BC (Canada)

    2008-07-01

    Remote communities may benefit from using tidal energy in terms of reduced diesel fuel consumption and the associated greenhouse gas emissions. A study was conducted to assess the feasibility for a tidal demonstration project on the Haida Gwaii, Queen Charlotte Islands. Candidate communities were scanned for resource potential, load profile, infrastructure distribution and community interest. This presentation focused on choosing an appropriate site for a given tidal power technology. Three hotspots in Masset Sound were identified as well as one hotspot at Juskatla Narrows. Technology providers were solicited for information on unit performance, cost, and trials to date. The presentation noted that demonstration or future commercial deployment is limited by resource and by the ability of the grid to accommodate tidal power. The presentation concluded with next steps which include publishing the study. tabs., figs.

  17. Rocky Flats Plant precipitate sludge surrogate vitrification demonstration. Technical Task Plan

    International Nuclear Information System (INIS)

    Cicero, C.A.; Bickford, D.F.; Bennert, D.M.; Overcamp, T.J.

    1994-01-01

    Technologies are being developed by the US Department of Energy's (DOE) Nuclear Facility sites to convert hazardous and mixed wastes to a form suitable for permanent disposal. The preferred disposal method would be one that is capable of consistently producing a durable leach resistant wasteform, while simultaneously minimizing disposal volumes. Vitrification, which has been declared the Best Demonstrated Available Technology (BDAT) for high-level radioactive waste disposal by the EPA, is capable of producing a highly durable wasteform that minimizes disposal volumes through organic destruction, moisture evaporation, and porosity reduction. However, this technology must be demonstrated over a range of waste characteristics, including compositions, chemistries, moistures, and physical characteristics to ensure that it is suitable for hazardous and mixed waste treatment

  18. Overview of seismic resistant design of Indian Nuclear Power Plants

    International Nuclear Information System (INIS)

    Sharma, G.K.; Hawaldar, R.V.K.P.; Vinod Kumar

    2007-01-01

    Safe operation of a Nuclear Power Plant (NPP) is of utmost importance. NPPs consist of various Structure, System and Equipment (SS and E) that are designed to resist the forces generated due to a natural phenomenon like earthquake. An earthquake causes severe oscillatory ground motion of short duration. Seismic resistant design of SS and E calls for evaluation of effect of severe ground shaking for assuring the structural integrity and operability during and after the occurrence of earthquake event. Overall exercise is a multi-disciplinary approach. First of standardized 220 MWe design reactor is Narora Atomic Power Station. Seismic design was carried out as per state of art then, for the first time. The twelve 220 MWe reactors and two 540 MWe reactors designed since 1975 have been seismically qualified for the earthquake loads expected in the region. Seismic design of 700 MWe reactor is under advanced stage of finalization. Seismic re-evaluation of six numbers of old plants has been completed as per latest state of art. Over the years, expertise have been developed at Nuclear Power Corporation of India Limited, Bhabha Atomic Research Centre, prominent educational institutes, research laboratories and engineering consultants in the country in the area of seismic design, analysis and shake table testing. (author)

  19. Electrical design requirements for electrode boilers for nuclear plants

    International Nuclear Information System (INIS)

    Kempker, M.J.

    1979-01-01

    Medium-voltage steam electrode boilers, in the 20- to 50-MW range, have become an attractive alternative to comparable fossil-fueled boilers as a source of auxiliary steam during the startup and normal shutdown of nuclear power plants. The electrode boiler represents a favorable option because of environmental, fire protection, and licensing considerations. However, this electrical option brings some difficult design problems for which solutions are required in order to integrate the electrode boiler into the plant low resistance grounded power system. These considerations include the effects of an unbalanced electrode boiler on the performance of polyphase induction motors, boiler grounding for personnel safety, boiler neutral grounding, and ground relaying

  20. Wind Power Plants Fundamentals, Design, Construction and Operation

    CERN Document Server

    Twele, Jochen

    2012-01-01

    Wind power plants teaches the physical foundations of usage of Wind Power. It includes the areas like Construction of Wind Power Plants, Design, Development of Production Series, Control, and discusses the dynamic forces acting on the systems as well as the power conversion and its connection to the distribution system. The book is written for graduate students, practitioners and inquisitive readers of any kind. It is based on lectures held at several universities. Its German version it already is the standard text book for courses on Wind Energy Engineering but serves also as reference for practising engineers.