A disposal centre for irradiated nuclear fuel: conceptual design study

International Nuclear Information System (INIS)

1980-09-01

This report describes a conceptual design of a disposal centre for irradiated nuclear fuel. The surface facilities consist of plants for the preparation of steel cylinders containing irradiated nuclear fuel immobilized in lead, shaft headframe buildings, and all necessary support facilities. The undergound disposal vault is located on one level at a depth of 1000 metres. The cylinders containing the irradiated fuel are emplaced on a one-metre thick layer of backfill material and then completely covered with backfill. All surface and subsurface facilities are described, operations and schedules are summarized, and cost estimates and manpower requirements are given. (auth)

Integral system for computer aided design of power line transmission towers; Sistema integral para el diseno asistido por computadora de torres de transmision

Energy Technology Data Exchange (ETDEWEB)

Nagore, Gabriel [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1987-12-31

The profitable use of the advantages offered by the computer aided design, through the integral system for the structural design of power transmission towers (SIDETT) will contribute to facilitate and improve the design of the structures that link the electric power generation centers with the consumers. [Espanol] El aprovechamiento de las ventajas que ofrece el diseno asistido por computadora, a traves del sistema integral para el diseno estructural de torres de transmision (SIDETT), contribuira a agilizar y mejorar el diseno de las estructuras que enlazan los centros de generacion de energia electrica con los consumidores.

Integral system for computer aided design of power line transmission towers; Sistema integral para el diseno asistido por computadora de torres de transmision

Energy Technology Data Exchange (ETDEWEB)

Nagore, Gabriel [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1986-12-31

The profitable use of the advantages offered by the computer aided design, through the integral system for the structural design of power transmission towers (SIDETT) will contribute to facilitate and improve the design of the structures that link the electric power generation centers with the consumers. [Espanol] El aprovechamiento de las ventajas que ofrece el diseno asistido por computadora, a traves del sistema integral para el diseno estructural de torres de transmision (SIDETT), contribuira a agilizar y mejorar el diseno de las estructuras que enlazan los centros de generacion de energia electrica con los consumidores.

Design study of advanced nuclear fuel recycle system. Conceptual study of recycle system using molten salt

International Nuclear Information System (INIS)

Kakehi, I.; Shirai, N.; Hatano, M.; Kajitani, M.; Yonezawa, S.; Kawai, T.; Kawamura, F.; Tobe, K.; Takahashi, K.

1996-12-01

For the purpose of developing the future nuclear fuel recycle system, the design study of the advanced nuclear fuel recycle system is being conducted. This report describes intermediate accomplishments in the conceptual system study of the advanced nuclear fuel recycle system. Fundamental concepts of this system is the recycle system using molten salt which intend to break through the conventional concepts of purex and pellet fuel system. Contents of studies in this period are as follows, 1)feasibility study of the process by Cd-cathode for nitride fuel, 2)application study for the molten salt of low melting point (AlCl3+organic salt), 3)research for decladding (advantage of decladding by heat treatment), 4)behavior of FPs in electrorefining (behavior of iodine and volatile FP chlorides, FPs behavior in chlorination), 5)criticality analysis in electrorefiner, 6)drawing of off-gas flow diagram, 7)drawing of process machinery concept (cathode processor, vibration packing), 8)evaluation for the amounts of the high level radioactive wastes, 9)quality of the recycle fuels (FPs contamination of recycle fuel), 10)conceptual study of in-cell handling system, 11)meaning of the advanced nuclear fuel recycle system. The conceptual system study will be completed in describing concepts of the system and discussing issues for the developments. (author)

CRBR nuclear, thermofluid, and advanced fuel conceptual design

International Nuclear Information System (INIS)

Dickson, P.W.

1975-01-01

The improvements effected in flow orificing and fuel conceptual design to achieve both the breeding ratio and fuel lifetime goals within the restrictions imposed upon the core and blanket are discussed. The effect of cladding temperature on fuel lifetime is illustrated for either inelastic strain limits or life fraction damage function limits. The temperature varies through life differently for different assemblies. The maximum cladding midwall temperature for the assembly illustrated is just over 1300 0 F at the beginning of life, also calculated on a conservative basis. This results in a lifetime of 80 MWD/Kg. An initial temperature of closer to 1230 0 F would be required to achieve a burnup capability of 150 MWD/Kg. It is thus apparent that either the temperatures of the cladding must be decreased, or improved cladding material is required in order to achieve 150 MWD/Kg. (auth)

Cost-benefit of energy saving in bioclimatic designs; Costo-beneficio del ahorro de energia en los disenos bioclimatios

Energy Technology Data Exchange (ETDEWEB)

Lopez P, J. Manuel A. [Doble dos soluciones de Negocios (Mexico); Morillon G, David; Rdriguez V, Luis [Universidad Autonoma de Mexico (Mexico)

2001-06-01

Tools and criteria for Cost-Benefit Analysis (ACB) of energy saving in bioclimatic designs (DB) are presented, for which a reference project is established. Based on the traditional design of air conditioning of a building and with the support of this one, the estimation of costs and benefits of the same building is made; but with passive systems. The tools used consider related resources such as time and money. Criteria are used such as: Present value (VP), Equivalent Annual Cost (EAC) and the Cost of the conserved Energy (CEC). The costs related to the construction, maintenance and operation of the design are taken into consideration. The differences between the design of reference and the Bioclimatic Design, established from the application of these criteria, allow evaluating the economic margin as far as operation and maintenance. Finally, the cost of the conserved energy of the bioclimatic measures is evaluated as an initiative for energy saving and calculates the present value from them in all of the period of useful life of the bioclimatic design. [Spanish] Se presentan herramientas y criterios para el analisis Costo-Beneficio (ACB) del ahorro de energia en los disenos bioclimaticos (DB), para lo cual se establece un proyecto de referencia. Basado en el diseno tradicional de la climatizacion de un edificio y con apoyo en este, se realiza la estimacion de costos y beneficios del mismo edificio; pero con sistema pasivos. Las herramientas usadas consideran recursos relacionados como son tiempo y dinero. Se usan criterios tales como: Valor Presente (VP), Costo Anual Equivalente (CAE) y el Costo de la Energia conservada (CEC). Se toman en consideracion los costos relacionados con la construccion, mantenimiento y operacion del diseno. Las diferencias entre el diseno de referencia y el Diseno Bioclimatico, establecidas a partir de la aplicacion de estos criterios, permiten evaluar el margen economico en cuanto a la operacion y el mantenimiento. Finalmente, se

Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

International Nuclear Information System (INIS)

Kim, S. H.; Koo, G. H.

2012-01-01

The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

Assessment of LMFBR spent fuel shipping cask concepts for the CRBRP and the US conceptual design study

International Nuclear Information System (INIS)

Pope, R.B.; Ortman, J.M.; Eakes, R.G.; Leisher, W.B.; Dupree, S.A.

1980-01-01

Study of conceptual shipping systems for CRBRP and CDS spent fuel has shown that systems significantly different from those used for LWR spent fuel will be required. In the conceptual design, liquid sodium was assumed to be the coolant in canisters containing the spent fuel assemblies, and multiple levels of containment were provided by canisters, an inner cask lid and an outer cask lid. Cask cooling at the reactor site during loading, and cooldown at the receiving site prior to unloading are significant but tractable problems

Spent nuclear fuel canister storage building conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

1996-01-01

This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

Spent nuclear fuel canister storage building conceptual design report

International Nuclear Information System (INIS)

Swenson, C.E.

1996-01-01

This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ''Technical Baseline and Updated Cost Estimate for the Canister Storage Building'', dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995

Application of the Pinch analysis for the design of a cogeneration system in a paper mill; Aplicacion del analisis Pinch para el diseno de un sistema de cogeneracion en una industria papelera

Energy Technology Data Exchange (ETDEWEB)

Mani Gonzalez, A. G.; Arriola Medellin, A. [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1997-12-31

The Pinch Analysis is a set of principles, tools and rules for the design that allow the engineer find the best way to configure the elements of a process. In the last ten years it has been utilized for the design of new processes as well as in the energy optimization of existing processes. In this paper the tools utilized for the integration of a cogeneration system in a process for the production of paper is presented. It is also presented how the combined treatment of the Pinch Analysis and the exergy concept allows to define, before the detailed design, the cogeneration potential, the fuel consumption and the amount of pollutant emissions for different cogeneration schemes. [Espanol] El analisis Pinch es un conjunto de principios, herramientas y reglas de diseno que permiten al ingeniero encontrar la mejor manera de configurar los elementos de un proceso. En los ultimos diez anos se ha utilizado para el diseno de procesos nuevos asi como en la optimacion energetica de procesos existentes. En el presente articulo se presentan las herramientas utilizadas para la integracion de un sistema de cogeneracion en un proceso de produccion de papel. Se muestra tambien como el tratamiento combinado del analisis Pinch y el concepto de energia permite definir, antes del diseno detallado, el potencial de cogeneracion, el consumo de combustible y la cantidad de emisiones contaminantes para diferentes esquemas de cogeneracion.

Application of the Pinch analysis for the design of a cogeneration system in a paper mill; Aplicacion del analisis Pinch para el diseno de un sistema de cogeneracion en una industria papelera

Energy Technology Data Exchange (ETDEWEB)

Mani Gonzalez, A G; Arriola Medellin, A [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1998-12-31

The Pinch Analysis is a set of principles, tools and rules for the design that allow the engineer find the best way to configure the elements of a process. In the last ten years it has been utilized for the design of new processes as well as in the energy optimization of existing processes. In this paper the tools utilized for the integration of a cogeneration system in a process for the production of paper is presented. It is also presented how the combined treatment of the Pinch Analysis and the exergy concept allows to define, before the detailed design, the cogeneration potential, the fuel consumption and the amount of pollutant emissions for different cogeneration schemes. [Espanol] El analisis Pinch es un conjunto de principios, herramientas y reglas de diseno que permiten al ingeniero encontrar la mejor manera de configurar los elementos de un proceso. En los ultimos diez anos se ha utilizado para el diseno de procesos nuevos asi como en la optimacion energetica de procesos existentes. En el presente articulo se presentan las herramientas utilizadas para la integracion de un sistema de cogeneracion en un proceso de produccion de papel. Se muestra tambien como el tratamiento combinado del analisis Pinch y el concepto de energia permite definir, antes del diseno detallado, el potencial de cogeneracion, el consumo de combustible y la cantidad de emisiones contaminantes para diferentes esquemas de cogeneracion.

Conceptual design report for the ICPP spent nuclear fuel dry storage project

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-07-01

The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel storage rack

International Nuclear Information System (INIS)

Tonny Lanyau; Mohd Fazli Zakaria; Zaredah Hashim; Ahmad Nabil Ab Rahim; Mohammad Suhaimi Kassim

2010-01-01

PUSPATI TRIGA Reactor (RTP) is a pool type research reactor with 1MW thermal power. It has been safely operated since 28 June 1982. During 28 years of safe operation, there are several systems and components of the RTP that have been maintained, repaired, upgraded and replaced in order to maintain its function and safety conditions. RTP has been proposed to be upgraded so that optimum operation of RTP could be achieved as well as fulfill the future needs. Thus, competencies and technical capabilities were needed to design and develop the reactor system. In the meantime, there is system or component need to be maintained such as fuel elements. Since early operation, most of the fuel elements still can be used and none of the fuel elements was replaced or sent for reprocessing and final disposal. Towards the power upgrading, preparation of spent fuel storage is needed for temporary storing of the fuels discharged from the reactor core. The spent fuel storage rack will be located in the spent fuel pool to accommodate the spent fuels before it is send to reprocessing or final disposal. This paper proposes the conceptual design of the spent fuel storage rack. The output of this paper focused on the physical and engineering design of the spent fuel storage. (author)

The conceptual flowsheet of effluent treatment during preparing spherical fuel elements of HTR

Energy Technology Data Exchange (ETDEWEB)

Ying, Quan, E-mail: quanying@tsinghua.edu.cn; Xiao-tong, Chen; Bing, Liu; Gen-na, Fu; Yang, Wang; You-lin, Shao; Zhen-ming, Lu; Ya-ping, Tang; Chun-he, Tang

2014-05-01

High temperature gas-cooled reactor (HTR) is one of the advanced nuclear reactors owing to its inherent safety and broad applications. For HTR, one of the key components is the ceramic fuel element. During the preparation of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was established. According to the above treatment process, the uranium concentration was decreased from 200 mg/l to the level of discharged standard.

Revised conceptual designs for the FMDP MOX fresh fuel transport package

International Nuclear Information System (INIS)

Ludwig, S.B.; Michelhaugh, R.D.; Shappert, L.B.; Chae, S.M.; Tang, J.S.

1998-03-01

The revised conceptual designs described in this document provide a foundation for the development and certification of final transport package designs that will be needed to support the disposition of surplus weapons-grade plutonium as mixed-oxide (MOX) fuel in commercial light-water reactors in the US. This document is intended to describe the revised package design concepts and summarize the results of preliminary analyses and assessments of two new concepts for fresh MOX fuel transport packages that have been developed by Oak Ridge National Laboratory during the past year in support of the Department of Energy/Office of Fissile Materials Disposition

Conceptual design and cost estimation of dry cask storage facility for spent fuel

International Nuclear Information System (INIS)

Maki, Yasuro; Hironaga, Michihiko; Kitano, Koichi; Shidahara, Isao; Shiomi, Satoshi; Ohnuma, Hiroshi; Saegusa, Toshiari

1985-01-01

In order to propose an optimum storage method of spent fuel, studies on the technical and economical evaluation of various storage methods have been carried out. This report is one of the results of the study and deals with storage facility of dry cask storage. The basic condition of this work conforms to ''Basic Condition for Spent Fuel Storage'' prepared by Project Group of Spent Fuel Dry Storage at July 1984. Concerning the structural system of cask storage facilities, trench structure system and concrete silo system are selected for storage at reactor (AR), and a reinforced concrete structure of simple design and a structure with membrance roof are selected for away from reactor (AFR) storage. The basic thinking of this selection are (1) cask is put charge of safety against to radioactivity and (2) storage facility is simplified. Conceptual designs are made for the selected storage facilities according to the basic condition. Attached facilities of storage yard structure (these are cask handling facility, cask supervising facility, cask maintenance facility, radioactivity control facility, damaged fuel inspection and repack facility, waste management facility) are also designed. Cost estimation of cask storage facility are made on the basis of the conceptual design. (author)

The conceptual design of the standard and the reduced fuel assemblies for an advanced research reactor

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Cho, Yeong Garp; Yoon, Doo Byung; Dan, Ho Jin; Chae, Hee Tack; Park, Cheol

2005-01-01

HANARO (Hi-flux Advanced Neutron Application Reactor), is an open-tank-in-pool type research reactor with a thermal power of 30MW. The HANARO has been operating at Korea Atomic Energy Research Institute since 1995. Based on the technical experiences in design and operation for the HANARO, the design of an Advanced Research Reactor (ARR) was launched by KAERI in 2002. The final goal of the project is to develop a new and advanced research reactor model which is superior in safety and economical aspects. This paper summarizes the design improvements of the conceptually designed standard fuel assembly based on the analysis results for the nuclear physics. It includes also the design of the reduced fuel assembly in conjunction with the flow tube as the fuel channel and the guide of the absorber rod. In the near future, the feasibility of the conceptually designed fuel assemblies of the ARR will be verified by investigating the dynamic and the thermal behaviors of the fuel assembly submerged in coolant

Conceptual design study of a concrete canister spent-fuel storage facility

International Nuclear Information System (INIS)

Lidfors, E.D.; Tabe, T.; Johnson, H.M.

1979-01-01

This report presents a conceptual design study for the interim storage of CANDU spent fuel in concrete canisters. The canisters will be concrete flasks, which contain fuel prepackaged in double steel containment, and will be cooled by natural air convection. This is one of the methods proposed as a potential alternative to water pool storage. A preliminary study of this concept was done by CAFS (Committee Assessing Fuel Storage), and WNRE (Whiteshell Nuclear Research Establishment) is currently conducting a development and demonstration program. This study of a central facility for the storage of all Canadian spent fuel arisings to the year 2000 was completed in 1975. A brief description of the facilities required and the operations involved, a summary of costs, a survey of the monitoring requirements and a prediction of the personnel exposures associated with this method of storing spent fuel are reported here. The estimated total cost of interim storage in cylindrical canisters at a central site is $6.02/kg U (1975 dollars). Approximately half of this cost is incurred in the shipment of fuel from the reactors to the storage facility. (author)

Design optimization of the Laguna Verde nuclear power station fuel recharge; Optimacion del diseno de recargas de combustible para la Central Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Cortes Campos, Carlos Cristobal [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico)

1991-12-31

It is described, in general terms, the procedure that is followed to accomplish the optimization of the recharge design, and an example is shown where this procedure was applied for the analysis of the type BWR reactor of Unit No. 1 of the Laguna Verde Nuclear Power Station. [Espanol] Se describe en terminos generales el procedimiento que se sigue para realizar la optimacion del diseno de recargas, y se muestra un ejemplo en el que se utilizo dicho procedimiento para el analisis del reactor tipo BWR de la unidad 1, de la Central Laguna Verde (CLV).

Design optimization of the Laguna Verde nuclear power station fuel recharge; Optimacion del diseno de recargas de combustible para la Central Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Cortes Campos, Carlos Cristobal [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico)

1992-12-31

It is described, in general terms, the procedure that is followed to accomplish the optimization of the recharge design, and an example is shown where this procedure was applied for the analysis of the type BWR reactor of Unit No. 1 of the Laguna Verde Nuclear Power Station. [Espanol] Se describe en terminos generales el procedimiento que se sigue para realizar la optimacion del diseno de recargas, y se muestra un ejemplo en el que se utilizo dicho procedimiento para el analisis del reactor tipo BWR de la unidad 1, de la Central Laguna Verde (CLV).

Application of core structural design guidelines in conceptual fuel pin design

International Nuclear Information System (INIS)

Patel, M.R.; Stephen, J.D.

1979-01-01

The paper describes an application of the Draft RDT Standards F9-7, -8, and -9 to conceptual design of Fast Breeder Reactor (FBR) fuel pins. The Standards are being developed to provide guidelines for structural analysis and design of the FBR core components which have limited ductility at high fluences and are not addressed by the prevalent codes. The development is guided by a national working group sponsored by the Division of Reactor Researcch and Technology of the Department of Energy. The development program summarized in the paper includes establishment of design margins consistent with the test data and component performance requirements, and application of the design rules in various design activities. The application program insures that the quantities required for proper application of the design rules are available from the analysis methods and test data, and that the use of the same design rules in different analysis tools used at different stages of a component design producees consistent results. This is illustrated in the paper by application of the design rules in the analysis methods developed for conceptual and more detailed designs of an FBR fuel pin

Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel pool cooling system

International Nuclear Information System (INIS)

Tonny Lanyau; Mazleha Maskin; Mohd Fazli Zakaria; Mohmammad Suhaimi Kassim; Ahmad Nabil Abdul Rahim; Phongsakorn Prak Tom; Mohd Fairus Abdul Farid; Mohd Huzair Hussain

2012-01-01

After undergo about 30 years of safe operation, Reactor TRIGA PUSPATI (RTP) was planned to be upgraded to ensure continuous operation at optimum safety condition. In the meantime, upgrading is essential to get higher flux to diversify the reactor utilization. Spent fuel pool is needed for temporary storage of the irradiated fuel before sending it back to original country for reprocessing, reuse after the upgrading accomplished or final disposal. The irradiated fuel elements need to be secure physically with continuous cooling to ensure the safety of the fuels itself. The decay heat probably still exist even though the fuel elements not in the reactor core. Therefore, appropriate cooling is required to remove the heat produced by decay of the fission product in the irradiated fuel element. The design of spent fuel pool cooling system (SFPCS) was come to mind in order to provide the sufficient cooling to the irradiated fuel elements and also as a shielding. The spent fuel pool cooling system generally equipped with pumps, heat exchanger, water storage tank, valve and piping. The design of the system is based on criteria of the primary cooling system. This paper provides the conceptual design of the spent fuel cooling system. (author)

Conceptual Design of Structural Components of a Dual Cooled Fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Kang-Hee; Kim, Jae-Yong; Yoon, Kyung-Ho

2008-01-15

A dual cooled fuel, featured by an internal as well as an external coolant flow passage of a fuel rod, was suggested to enable a large-scaled power-uprate of PWR plant and launched as one of the National Nuclear R and D Projects in 2007. It is necessary to make the dual cooled fuel be compatible with an OPR-1000 system to maximize the economy. Also, the structural components of the dual cooled fuel should be designed to realize their features. To this end, a conceptual design of a spacer grid, outer and center guide tubes, and top and bottom end pieces has been carried out in the project 'Development of Design Technology for Dual Cooled Fuel Structure'. For the spacer grids, it is suggested that springs and dimples are located at or near the cross points of the straps due to a considerably narrowed rod-to-rod gap. Candidate shapes of the grids were also developed and applied for domestic patents. For the outer and center guide tubes, a dual tube like a fuel rod was suggested to make the subchannel areas around the guide tubes be similar to those around the fuel rods of enlarged diameter. It was applied for the domestic patent as well. For the top and bottom end pieces, the shape and pattern have been changed from the conventional ones reflecting the fuel rods' changes. Technical issues and method of resolution for each components were listed up for a basic design works in the following years.

System for the design of electrical distribution substations; Sistema para el diseno de subestaciones electricas de distribucion

Energy Technology Data Exchange (ETDEWEB)

Zayas Perez, Benjamin Eddie; Islas Perez, Eduardo; Bahena Rada, Jessica Liliana; Romero Lima, Jesus [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico); Sierra Rodriguez, Benjamin; Moreno Diaz, Humberto [Comision Federal de Electricidad (Mexico)

2011-07-01

The physical design of electric substations involves a variety of activities, such as developing, reviewing and performing modifications to different design proposals, estimating costs and producing documents to be integrated into an executive project. These activities are usually conducted at the Distribution Divisions of the Federal Electric Commission (CFE) through two separate processes: the development of engineering plans with computer-aided design software (CAD) and volumetric calculations for the initial budget using conventional spreadsheets. The documents are manually developed on an individual basis, without the existence of any functions to allow for automatically updating the design modifications of any of the plans. The corresponding initial budget becomes a lengthy and laborious process with the likelihood of inconsistencies among the documents. In order to optimize the design process for the public and electromechanical works of electric substations (ES) in a three-dimensional (3D) graphic environment, the IIE's Management of Process Supervision, in coordination with the CFE Distribution Subdivision, developed the System for the Design of Electric Substation Distribution (SDESD). This system makes use of the advantages of CAD3D technology to develop detail engineering, cost engineering to calculate initial budgets and virtual reality for three-dimensional interactive navigation. It is worth mentioning that this system is currently in the implementation phase at the CFE Distribution Subdivision. This article briefly describes the main stages in the development of the project, as well as the functioning and principal components of the system. [Spanish] El diseno fisico de subestaciones electricas implica una variedad de actividades, tales como elaborar, revisar y efectuar modificaciones de diferentes propuestas de diseno, estimar costos y producir documentos, para integrarlos como parte de un proyecto ejecutivo. Estas actividades usualmente se

Conceptual structure design of experimental facility for advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Joo, J. S.; Koo, J. H.; Jung, W. M.; Jo, I. J.; Kook, D. H.; Yoo, K. S.

2003-01-01

A study on the advanced spent fuel conditioning process (ACP) is carring out for the effective management of spent fuels of domestic nuclear power plants. This study presents basic shielding design, modification of IMEF's reserve hot cell facility which reserved for future usage, conceptual and structural architecture design of ACP hot cell and its contents, etc. considering the characteristics of ACP. The results of this study will be used for the basic and detail design of ACP demonstration facility, and utilized as basic data for the safety evaluation as essential data for the licensing of the ACP facility

Conceptual design of experimental LFR fuel element for testing in TRIGA reactor, ACPR zone

International Nuclear Information System (INIS)

Nastase, D.; Olteanu, G.; Ioan, M.; Pauna, E.

2013-01-01

In the pulsed area of the TRIGA reactor (ACPR zone), the irradiation tests called ''rapid insertions of reactivity on different types of nuclear fuel elements'' are usually realized. During these tests, in the fuel element high powers for a relatively short period of time (about few milliseconds) are generated. The generated heat in fuel pellets raise their central temperature to values over 100 deg C. The conceptual design of an experimental fuel element proposed to be developed and presented in this paper must fulfill a couple of requirements, as follows: to ensure full compatibility with irradiation device sample holder (compatibility is achieved through reduced length of the fuel stack pellets - this way assures a flow flattening on the entire length of the fuel element); to be compatible with the project of irradiated fuel bundle in Lead cooled Fast Reactors (LFR). (authors)

Conceptual Assessment of a Fresh Fuel Transport Package for KJRR

Energy Technology Data Exchange (ETDEWEB)

Lee, Ju-Chan; Choi, W. S.; Bang, K. S.; Yu, S. H.; Park, J. S.; Yang, Y. Y. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The IAEA and domestic regulations stipulate that the fissile material transport package be subjected to the cumulative effects of a 9 m drop, 1 m puncture, 800 ℃ thermal and water leakage tests. A fissile material transport package should be maintained the subcriticality during the normal and accident conditions for contingency of leakage of water into or out of package, rearrangement of the contents, reduction of spaces and temperature changes. KAERI has been developing a fresh fuel transport package for Kijang research reactor (KJRR). This paper describes a conceptual design and preliminary safety analysis of the transport package for KJRR. The transport package was designed for shipment of a fresh fuel and a FM (Fission Molybdenum) target. Low-enriched uranium (LEU) of U-Mo fuel with U-235 enrichment of 19.75 w/o is used as a research reactor fuel. And LEU of UAlx-Al with U-235 enrichment of 19.75 w/o is used as a FM target material. The transport package was designed for shipment of a fresh fuel and a FM target. Safety analyses were conducted on all areas, including criticality, structural, and thermal fields. In the criticality analysis, effective neutron multiplication factors were below the criticality safety limit. In the structural analysis, the maximum stress satisfied the stress requirement stipulated in the ASME code. After 9 m free drop and 1 m puncture test, there was no significant deformation of fuel basket to cause a criticality. In the thermal analysis, the maximum temperatures at each part were lower than the allowable values.

Bioclimatic analysis and its impact within the design methodology; Analisis bioclimatico y su impacto dentro de la metodologia de diseno

Energy Technology Data Exchange (ETDEWEB)

Fuentes Fleixanet, Victor A; Rodriguez Viqueira Manuel [Universidad Autonoma Metropolitana Azcapotzalco, Mexico, D.F(Mexico)

2000-07-01

This paper is a proposal for the systematization of the process of bioclimatic design, particularly in its analytical stage. It is based on the proposals of the classic researchers such as: Olgyay, Givoni and Szokolay, but also in new contributions like those of Yeang. Nevertheless, it is about a methodology adapted to the specific requirements of teaching and research that are carried out in the Bioclimatic Design Laboratory of the Universidad Autonoma Metropolitana, Azcapotzalco Unit. What is presented here is a part of a very ample research project that intends to develop or to structure a design methodology that incorporates the environmental variables related to man and architecture: of course, with the objective of facilitating the designer work so that the final architectonic product has a more harmonic response to the environment, offers conditions of integral well- being to its occupants and be is efficient in the handling of the resources. It also looks for the definition of the necessary analysis, design and evaluation tools, necessary in order that the application of the methodology is clear, simple and can be carried out by any designer, even if he does not have a wide knowledge of bioclimatic design. [Spanish] Este trabajo es una propuesta para la sistematizacion del proceso de diseno bioclimatico, particularmente en su etapa analitica. Se basa en las propuestas de los investigadores clasicos como: Olgyay, Givoni y Szokolay, pero tambien en nuevas aportaciones como las de Yeang. Sin embargo se trata de una metodologia adaptada a los requerimientos especificos de docencia e investigacion que se llevan a cabo en el Laboratorio de Diseno Bioclimatico de la Universidad Autonoma Metropolitana Unidad Azcapotzalco. Lo que aqui se presenta forma parte de un proyecto de investigacion muy amplio que pretende, desarrollar o estructurar una metodologia de diseno que incorpore las variables ambientales relacionadas con el hombre y la arquitectura: desde luego, con

Conceptual design of coal-fueled diesel system for stationary power applications

Energy Technology Data Exchange (ETDEWEB)

1989-05-01

A preliminary conceptual design of a coal-fueled diesel system was prepared as part of a previous systems study. Since then, our team has accumulated extensive results from testing coal-water slurry on the 13-inch bore JS engine (400 rpm) in 1987 and 1988. These results provided new insights into preferred design concepts for engine components. One objective, therefore, was to revise the preliminary design to incorporate these preferred design concepts. In addition there were certain areas where additional, more detailed analysis was required as a result of the previous conceptual design. Another objective, therefore was to perform additional detailed design efforts, such as: (1) market applications and engine sizes, (2) coal-water slurry cleaning and grinding processes, (3) emission controls and hot gas contaminant controls, (4) component durability, (5) cost and performance assessments. (VC)

Performance and fuel cycle cost analysis of one Janus 30 conceptual design for several fuel element design options

Energy Technology Data Exchange (ETDEWEB)

Nurdin, Martias [Research Centre for Nuclear Techniques, National Atomic Energy Agency (Indonesia); Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

The performance and fuel cycle costs for a 25 MW, JANUS 30 reactor conceptual design by INTERATOM, Federal Republic of Germany, for BATAN, Republic of Indonesia have been studied using 19.75% enriched uranium in four fuel element design options. All of these fuel element designs have either been proposed by INTERATOM for various reactors or are currently in use with 93% enriched uranium in reactors in the Federal Republic of Germany. Aluminide, oxide, and silicide fuels were studied for selected designs using the range of uranium densities that are either currently qualified or are being developed and demonstrated internationally. These uranium densities include 1.7-2.3 g/cm{sup 3} in aluminide fuel, 1.7-3.2 g/cm{sup 3} in oxide fuel, and 2.9-6.8 g/cm{sup 3} in silicide fuel. As of November 1982) both the aluminide and the oxide fuels with about 1.7 g U/cm{sup 3} are considered to be fully-proven for licensing purposes. Irradiation screening and proof testing of fuels with uranium densities greater than 1.7 g/cm{sup 3} are currently in progress, and these tests need to be completed in order to obtain licensing authorization for routine reactor use. To assess the long-term fuel adaptation strategy as well as the present fuel acceptance, reactor performance and annual fuel cycle costs were computed for seventeen cases based on a representative end-of-cycle excess reactivity and duty factor. In addition, a study was made to provide data for evaluating the trade-off between the increased safety associated with thicker cladding and the economic penalty due to increased fuel consumption. (author)

Conceptual assessment and thermal hydraulic analysis of MVDS system for the dry storage of reduced metal fuel

International Nuclear Information System (INIS)

Lee, J. C.; Bang, K. S.; Shin, H. S.; Joo, J. S.; Su, K. S.; Kim, H. D.

2003-01-01

Conceptual assessment and thermal hydraulic analysis of MVDS storage system have been carried out for application of reduced metal fuel. The storage concept was established considering the optimum weight, storage volume and thermal efficiency. The capacity of MVDS system for loading the reduced metal fuel has four times as compared with existing PWR fuel storage system. In the results of thermal analysis, the maximum temperature of metal fuel was estimated to be 110 .deg. C which is lower than the allowable value under normal operation condition. Therefore, it is shown that the MVDS system can feasibly accomodate the reduced metal fuel in aspect of thermal safety

The Conceptual Design for Tubular Fuel Assemblies of an Advanced Research Reactor

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Dan, Ho Jin; Cho, Yeong Garp; Yoon, Doo Byung; Park, Cheol

2005-05-01

An Advanced Research Reactor(ARR) is being designed by KAERI since 2002. The final goal of the project is to develop a new and unique research reactor model which is superior in safety and economical aspects. In this work, the conceptual design for tubular fuel assemblies was carried out to enhance the previous model. The shape optimization of the cross section of the top guide was performed, and the swaging procedure in connecting fuel plates and stiffeners was developed. Moreover to reflect changes in number and size of fuel plates, related parts of the standard and the reduced fuel assemblies were redesigned. The top guide should suppress the vibration of the fuel assembly due to coolant and resist against material failures owing to fatigue and yield. In order to gain these design requirements, we have optimized the section profile of the top guide. To confirm manufacturing aspects, the swaging procedure was developed and its performance was tested. The results of tangential tensile test and axial compression test guaranteed that the fixing state between fuel plates and stiffeners is firm enough to hold each other. In addition, due to changes in number and size of fuel plates, the outer cross section of the fuel assembly was expanded and the diameter of the spacer tube was reduced. Reflecting these design changes, top/bottom guide, top guide cover, spring, spring cover, and receptacle were readjusted. Based on the technical experiences on the design and operation of the HANARO, the standard and the reduced fuel assemblies will be verified by performing various tests and analysis

Conceptual design of an interim dry storage system for the Atucha nuclear power plant spent fuels

International Nuclear Information System (INIS)

Nassini, Horacio E.P.; Fuenzalida Troyano, C.S.; Bevilacqua, Arturo M.; Bergallo, Juan E.

2005-01-01

The Atucha I nuclear power station, after completing the rearrangement and consolidation of the spent fuels in the two existing interim wet storage pools, will have enough room for the storage of spent fuel from the operation of the reactor till December 2014. If the operation is extended beyond 2014, or if the reactor is decommissioned, it will be necessary to empty both pools and to transfer the spent fuels to a dry storage facility. This paper shows the progress achieved in the conceptual design of a dry storage system for Atucha I spent fuels, which also has to be adequate, without modifications, for the storage of fuels from the second unity of the nuclear power station, Atucha II, that is now under construction. (author) [es

Hybrid 95: Software for the design of hybrid systems, Potrerillo of Santa Teresa; Hybrid 95: Software de diseno de sistemas hibridos, Potrerillo de Santa Teresa

Energy Technology Data Exchange (ETDEWEB)

Casaravilla, G.; Chaer, R. [Instituto de Ingenieria Electrica, Facultad de Ingenieria, Universidad de la Republica, Montevideo (Uruguay)

1997-12-31

This paper describes the software Hybrid 95 and how it is used for the design of hybrid systems. As an application example some of the design stages are looked over (determination of the wind resource, simulations for sizing generation/storage and topology) of the Potrerillo installation. Taking into account the site clear ecological connotations, during the design consideration is made of the visual impact that represents the possible wind-generator, therefore several locations and wind-generator heights are studied, even arriving to the wind-generation elimination. [Espanol] Este trabajo describe el software Hybrid 95 y como se utiliza a los efectos del diseno de sistemas hibridos. Como ejemplo de aplicacion se recorren algunas de las etapas del diseno (determinacion del recurso eolico, simulaciones para el dimensionado de generacion/acumulacion y topologia) de la instalacion de Potrerillo. Teniendo en cuenta las claras connotaciones ecologistas del lugar, durante el diseno se tiene presente el impacto visual que representa el eventual aerogenerador por lo que se estudian diversas ubicaciones y alturas de aerogenerador, llegandose incluso a estudiar la eliminacion de la generacion eolica.

Hybrid 95: Software for the design of hybrid systems, Potrerillo of Santa Teresa; Hybrid 95: Software de diseno de sistemas hibridos, Potrerillo de Santa Teresa

Energy Technology Data Exchange (ETDEWEB)

Casaravilla, G; Chaer, R [Instituto de Ingenieria Electrica, Facultad de Ingenieria, Universidad de la Republica, Montevideo (Uruguay)

1998-12-31

This paper describes the software Hybrid 95 and how it is used for the design of hybrid systems. As an application example some of the design stages are looked over (determination of the wind resource, simulations for sizing generation/storage and topology) of the Potrerillo installation. Taking into account the site clear ecological connotations, during the design consideration is made of the visual impact that represents the possible wind-generator, therefore several locations and wind-generator heights are studied, even arriving to the wind-generation elimination. [Espanol] Este trabajo describe el software Hybrid 95 y como se utiliza a los efectos del diseno de sistemas hibridos. Como ejemplo de aplicacion se recorren algunas de las etapas del diseno (determinacion del recurso eolico, simulaciones para el dimensionado de generacion/acumulacion y topologia) de la instalacion de Potrerillo. Teniendo en cuenta las claras connotaciones ecologistas del lugar, durante el diseno se tiene presente el impacto visual que representa el eventual aerogenerador por lo que se estudian diversas ubicaciones y alturas de aerogenerador, llegandose incluso a estudiar la eliminacion de la generacion eolica.

Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

International Nuclear Information System (INIS)

Lord, D.L.; Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L.

1993-04-01

A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed

Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

Energy Technology Data Exchange (ETDEWEB)

Lord, D.L. [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States); Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1993-04-01

A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed.

Feed-water heaters alternative design comparison; Comparacion de disenos alternativos de calentadores

Energy Technology Data Exchange (ETDEWEB)

Torres Toledano, Gerardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1988-12-31

A procedure is presented for the alternative design comparison of feed water heaters, based in the failure records of damaged tubes during operation. The procedure is used for cases in which non-continuous or random inspections are made to the feed-water heaters. [Espanol] Se presenta un procedimiento para comparar disenos alternativos de calentadores, basandose en los registros de fallas de los tubos rotos acumuladas durante su operacion. El procedimiento se emplea para casos en los que se realizan inspecciones a los calentadores no continuas, ya sea periodicas o al azar.

Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

International Nuclear Information System (INIS)

Pope, R.B.; Diggs, J.M.

1982-04-01

Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented

Evaluation of spent fuel properties from a conceptual PEACER core

International Nuclear Information System (INIS)

Lim, Jae Yong; Kim, Myung Hyun; Kim, Chang Hyo; Hwang, Il Soon

2003-01-01

In this paper, a new conceptual core design, PEACER was evaluated in aspect of core performance and spent fuel properties. The core shape is like a pancake to increase axial neutron leakage. Square lattice array was applied which was suitable to decrease the flow speed of Pb-Bi coolant. Although over 30% TRU produced by pyroprocessing was loaded in U-Zr metal fuel, the cycle length of 1 year was achieved and the relative assembly power peaking was less than 1.3. In order to confirm nuclear performance of PEACER core design, several performance indices were adopted and developed. Simple indices such as FIR and FG were used to evaluate fissile breeding. BCM, TG, SNS, and OR calculated by plutonium composition vectors were chosen to distinguish the competency of proliferation resistance. For the estimation of transmutation capability, D-value and extended effective fission half-life time(T EX ) were used. According to these indices, the PEACER core had the better performance compared with other conventional reactor cores although fissile breeding was not acquired

CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL

Energy Technology Data Exchange (ETDEWEB)

Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

2008-02-01

The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for

National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Volume I. Conceptual design report

International Nuclear Information System (INIS)

1978-12-01

In February 1976, the Energy Research and Development Administration (ERDA), now the Department of Energy (DOE), established a National Waste Terminal Storage (NWTS) program. As a part of this program, two parallel conceptual design efforts were initiated in January 1977. One was for deep geologic storage, in domed salt, of high level waste resulting from the reprocessing of spent fuel. The other was for deep geologic storage of unreprocessed spent fuel in bedded salt. These two concepts are identified as NWTS Repository 1 and Repository 2, respectively. Repository 2 (NWTSR2) is the concept which is covered by this Conceptual Design Report. Volume I of the conceptual design report contains the following information: physical description of the report; project purpose and justification; principal safety, fire, and health hazards; environmental impact considerations; quality assurance considerations; assessment of operational interfaces; assessment of research and development interfaces; project schedule; proposed method of accomplishment; summary cost estimate; and outline specifications. The conceptual design for Repository 2 was developed in sufficient detail to permit determination of scope, engineering feasibility, schedule, and cost estimates, all of which are necessary for planning and budgeting the project

Conceptual design report for a Direct Hydrogen Proton Exchange Membrane Fuel Cell for transportation application

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-05

This report presents the conceptual design for a Direct-Hydrogen-Fueled Proton Exchange Membrane (PEM) Fuel Cell System for transportation applications. The design is based on the initial selection of the Chrysler LH sedan as the target vehicle with a 50 kW (gross) PEM Fuel Cell Stack (FCS) as the primary power source, a battery-powered Load Leveling Unit (LLU) for surge power requirements, an on-board hydrogen storage subsystem containing high pressure gaseous storage, a Gas Management Subsystem (GMS) to manage the hydrogen and air supplies for the FCS, and electronic controllers to control the electrical system. The design process has been dedicated to the use of Design-to-Cost (DTC) principles. The Direct Hydrogen-Powered PEM Fuel Cell Stack Hybrid Vehicle (DPHV) system is designed to operate on the Federal Urban Driving Schedule (FUDS) and Hiway Cycles. These cycles have been used to evaluate the vehicle performance with regard to range and hydrogen usage. The major constraints for the DPHV vehicle are vehicle and battery weight, transparency of the power system and drive train to the user, equivalence of fuel and life cycle costs to conventional vehicles, and vehicle range. The energy and power requirements are derived by the capability of the DPHV system to achieve an acceleration from 0 to 60 MPH within 12 seconds, and the capability to achieve and maintain a speed of 55 MPH on a grade of seven percent. The conceptual design for the DPHV vehicle is shown in a figure. A detailed description of the Hydrogen Storage Subsystem is given in section 4. A detailed description of the FCS Subsystem and GMS is given in section 3. A detailed description of the LLU, selection of the LLU energy source, and the power controller designs is given in section 5.

Design of very high speed electric generators;Diseno de generadores electricos de muy alta velocidad de giro

Energy Technology Data Exchange (ETDEWEB)

Labollita, Santiago [Universidad Nacional de Cuyo, Instituto Balseiro, Centro Atomico Bariloche (Argentina)

2008-07-01

This work approaches the design process of an electric generator suitable for running efficiently at high speed, driven by a turbo shaft.The axial flux concept was used.For the mechanical design of the prototype, cooling capacity and mounting method were considered, looking for simplicity of the parts evolved. Neodymium-iron-boron permanent magnets were used as magnetic source.For the electrical design, a calculation tool was developed in order to predict the prototype electrical parameters and optimize its geometry.The goal was to obtain 1 kW of electric power at a speed of 100,000 rpm.The efficiency and electrical behaviour of the prototype were characterized at speeds between 2,000 rpm and 30,000 rpm and then the behaviour at the design condition was predicted by obtaining an equivalent electric circuit.The estimated load voltage was 237 V as well as an electrical efficiency of 95%.Eddy current effects were not recognized. Increase of the internal resistance and decree of inductance were observed while raising the electric frequency.Finally, an electronic system was developed in order to use the prototype as a c.c. motor. Global performance was measured according to different supply characteristic. An optimum supply voltage was found.A maximum efficiency of 63% was reached.;El presente trabajo aborda el proceso de diseno de un generador electrico apto para funcionar eficientemente a alta velocidad, impulsado por un turboeje portatil. Se utilizo el concepto de flujo axial.En el diseno mecanico del prototipo se considero la capacidad de refrigeracion y la forma de montaje particular del conjunto, buscando la simplicidad de sus partes.Se utilizaron imanes permanentes de Neodimio-Hierro-Boro como fuente magnetica. Para el diseno electrico se genero una herramienta de calculo que permitio estimar los parametros electricos del prototipo y optimizar su geometria. Se busco obtener 1kW de potencia electrica a una velocidad de 100.000rpm.Se caracterizo el rendimiento y el

Conceptual design report for the spent fuel management technology research and test (SMATER) facility

Energy Technology Data Exchange (ETDEWEB)

Park, S W; Ro, S G; Lee, J S; Min, D K; Shin, Y J [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-10-01

This study was intended to develop concept for a pilot-scale remote operation facility for longer term management of spent fuel and therefrom to provide technical requirement for later basic design of the facility. Main scope of work for the study was to revise the past (1990) conceptual design in functions, scale, hot cell layout etc. based on user requirements. Technical reference was made to the PKA facility in Germany, through collaboration with appropriate partner, to elaborate the design and requirements. The study was focused on establishing design criteria and conceptual design of the SMATER facility. The results of this study should be an essential and useful basis upon optimization for further work to basic design of the facility. (author). 17 figs., 12 tabs.

Environmental psychology in the bioclimatic design; La psicologia ambiental en el diseno bioclimatico

Energy Technology Data Exchange (ETDEWEB)

Cubillas Talamante, Lorena Gpe [Facultad de Arquitectura, Universidad Autonoma de Baja California, Mexicali, Baja California (Mexico)

2005-07-01

The bioclimatic design profits of the natural resources for the improvement of the construction quality as well as the user's life. The combination of these two concepts within the architectonic design generates an interesting duality to consider within the understanding of the user's physical and psychological environment and the obtaining of a correlation of reality and dreams. Environmental psychology is focused on the researching of the psychological effects of the architectonic environment on the human experience and behavior. Human behavior works under parameters established during its development. Through the habits and styles of life practiced in childhood those parameters establish themselves to later on obtain satisfiers in the search of the life quality of human beings. [Spanish] El diseno bioclimatico aprovecha los recursos naturales para el mejoramiento de la calidad de la construccion y vida del usuario. La combinacion de estos dos conceptos dentro del diseno arquitectonico, genera dualidad interesante de entender el medio ambiente fisico y psicologico de los usuarios y obtener una identificacion entre la realidad y los suenos. La psicologia ambiental se ha enfocado a investigar sobre la relacion de los efectos psicologicos del ambiente construido o arquitectonico sobre la conducta y la experiencia humana. La conducta humana trabaja bajo parametros establecidos durante el desarrollo del mismo, a traves de los habitos y estilos de vida desde su infancia van marcando esos parametros que despues sirven para lograr satisfactores en la busqueda de la calidad de vida de los seres humanos.

System programs design of motors; Sistema de programas de diseno de motores

Energy Technology Data Exchange (ETDEWEB)

Diaz Gonzalez Palomas, Oscar; Ciprian Avila, Fernando [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1988-12-31

This paper describes the objective of creating the program system for induction motors design SIPRODIMO, its scope, its general characteristics, its structure and the results obtained with its application, as well as the service capacity developed by the Motors Area of the Instituto de Investigaciones Elctricas. [Espanol] En este articulo se describe el objetivo de crear el sistema de programas de diseno de motores de induccion, Siprodimo, su alcance, sus caracteristicas generales, su estructura y los resultados obtenidos con su aplicacion, asi como la capacidad de servicio desarrollada por el area de motores, del Instituto de Investigaciones Electricas.

105-N Fuel Storage Basin dewatering conceptual plan

International Nuclear Information System (INIS)

Schilperoort, D.L.

1996-11-01

This conceptual plan discusses the processes that will be used for draining and disposing of water from the 105-N Fuel Storage Basin (N Basin), and includes a description of the activities to control surface contamination and potential high dose rates encountered during dewatering. The 105-N Fuel Storage Basin is located in the 100-N Area of the Hanford Site in Richland, Washington. The processes for water disposal include water filtration, water sampling and analysis, tanker loading and unloading, surface decontamination and sealing, and clean out and disposal of residual debris and sediments during final pumping to remove the N Basin water. Water disposal is critical for the deactivation of N Reactor. A Memorandum of Understanding (MOU) between the US Department of Energy (DOE) Environmental Restoration (ER) Program and DOE Waste Management (WM) Program establishes the 200 East Effluent Treatment Facility (ETF) as the final treatment and disposal site for N Basin water and identifies pre-treatment requirements. This MOU states that water delivery will be completed no later than October 31, 1996, and will require a revision due to the current de-watering schedule date. The current MOU requires four micron filtration prior to shipment to ETF. The MOU revision for delivery date extension seeks to have the filtration limit increased to five microns, which eliminates the need for a second filter system and simplifies dewatering. For the purposes of this plan, it will be assumed that five micron filtration will be used

The climatic change and the new demands in dwelling design; El cambio climatico y las nuevas exigencias en el diseno de la vivienda

Energy Technology Data Exchange (ETDEWEB)

Tardan Waltz, Jenny [Asociacion de Empresas para el Ahorro de la Energia en la Edificacion, A.C. (Mexico)

2007-10-15

In Mexico, the culture of energy saving began more than one decade ago, but the benefits are not yet concrete. The Mexican society, requires new designs of dwellings adapted to its needs and in addition modifies the present systems of construction that do not include criteria of design for the thermal comfort, the energy saving and the sustainability. It is urgent to reconsider the urbanism of the cities, as well as the promotion of the concept of sustainable development in all the branches of human activity. The design of dwellings must significantly integrate the energy and environmental components to reduce to the power consumption and the CO{sub 2} emissions. [Spanish] En Mexico, la cultura de ahorro de energia se inicio hace mas de una decada, pero los beneficios aun no son palpables. La sociedad mexicana, requiere de nuevos disenos de viviendas que se adapten a sus necesidades y que ademas modifiquen los sistemas de construccion actuales que no incluyen criterios de diseno para el confort termico, el ahorro de energia y la sustentabilidad. Es urgente reconsiderar el urbanismo de las ciudades, asi como la promocion del concepto de desarrollo sustentable en todas las ramas de la actividad humana. El diseno de la vivienda debe integrar los componentes energeticos y medioambientales, para reducir significativamente el consumo de energia y las emisiones de CO{sub 2}.

Safety Consideration for a Wet Interim Spent Fuel Store at Conceptual Design Stage

International Nuclear Information System (INIS)

Astoux, Marion

2014-01-01

EDF Energy plans to build and operate two UK EPRs at the Hinkley Point C (HPC) site in Somerset, England. Spent fuel from the UK EPRs will need to be managed from the time it is discharged from the reactor until it is ultimately disposed of and this will involve storing the spent fuel for a period in the fuel building and thereafter in a dedicated interim facility until it can be emplaced within the UK Geological Disposal Facility. EDF Energy has proposed that this interim store should be located on the Hinkley Point site which is consistent with UK policy. This Interim Spent Fuel Store (ISFS) will have the capability to store for at least one hundred years the spent fuel arising from the operation of the two EPR units (sixty years operation). Therefore, specificities regarding the lifetime of the facility have to be accounted for its design. The choice of interim storage technology was considered in some depth for the HPC project and wet storage (pool) was selected. The facility is currently at conceptual design stage, although its construction will be part of main site construction phase. Safety functions and safety requirements for this storage facility have been defined, in compliance with WENRA 'Waste and Spent Fuel Storage - Safety Reference Level Report' and IAEA Specific Safety Guide no. 15 'Storage of Spent Nuclear Fuel'. EDF technical know-how, operational feedback on existing storage pools, UK regulatory context and Fukushima experience feedback have also been accounted for. Achievement of the safety functions as passively as reasonably practicable is a key issue for the design, especially in accident situations. Regarding lifetime aspects, ageing management of equipments, optimisation of the refurbishment, climate change, passivity of the facility, and long-term achievement of the safety functions are among the subjects to consider. Adequate Operational Limits and Conditions will also have to be defined, to enable the long-term achievement of the safety

A SCWR core design with a conceptual fuel assembly using a cruciform moderator

International Nuclear Information System (INIS)

Bae, Kang Mok; Joo, Hyung Kook; Lee, Hyun Chul; Noh, Jae Man; Bae, Yoon Yong

2005-01-01

A super critical water cooled reactor (SCWR) system has a potential to compete with the advanced fossil plant by achieving a high thermal efficiency up to 44% and a plant simplification by eliminating steam generators, steam dryers, steam separators, and recirculation pumps. Due to these advantages, a SCWR is considered as one of the most promising nuclear plants for the Generation-IV (Gen-IV) system. As a first step of a feasibility study a rectangular fuel assembly with a cruciform solid moderator was suggested as a conceptual assembly design at the Korea Atomic Energy Research Institute (KAERI) for the SCWR on a thermal neutron spectrum. In this paper, based on the system parameters proposed by the Gen-IV road map, a preliminary SCWR core design was performed using a conceptual assembly design focused on the power shape control, reactivity coefficients, and cladding temperature limit

Fuel to Moderator Ratio Sensitivity Study Using Water Rod Moderator in SCWR Conceptual Core Design

International Nuclear Information System (INIS)

Bae, Seong Man; Kim, Yong Bae; Park, Dong Hwan; Lee, Kwang Ho

2009-01-01

The conceptual operating condition of Super Critical Water-cooled Reactor (SCWR) is above critical point of water, such that the coolant temperature ranges from 280 .deg. C to 510 .deg. C with a pressure of 25MPa. This operating condition makes an SCWR have both merits and demerits when compared with current Light Water Reactors (LWRs). One of the demerits of SCWR is degradation of neutron moderation due to a lower water density from ∼0.1g/cm 3 to ∼0.7g/cm 3 under a high coolant temperature condition. Therefore it is necessary to enhance the moderation capability for SCWR to slow down the fast fission neutrons. Many SCWR designs have a water rod concept as an additional moderator, because water has a good moderation capability. Through reviewing the previous water rod assembly designs, it was identified that a sensitivity study is required to optimize fuel assembly pitch to increase the neutron economy. In this paper, the results of the conceptual assembly design sensitivity study which focuses on the comparison of sensitivity for the fuel pitch to diameter (P/D) ratio using water rod moderator, are presented

Environmental psychology in the bioclimatic design; La psicologia ambiental en el diseno bioclimatico

Energy Technology Data Exchange (ETDEWEB)

Cubillas Talamante, Lorena Gpe [Facultad de Arquitectura, Universidad Autonoma de Baja California, Mexicali, Baja California (Mexico)

2005-07-01

The bioclimatic design profits of the natural resources for the improvement of the construction quality as well as the user's life. The combination of these two concepts within the architectonic design generates an interesting duality to consider within the understanding of the user's physical and psychological environment and the obtaining of a correlation of reality and dreams. Environmental psychology is focused on the researching of the psychological effects of the architectonic environment on the human experience and behavior. Human behavior works under parameters established during its development. Through the habits and styles of life practiced in childhood those parameters establish themselves to later on obtain satisfiers in the search of the life quality of human beings. [Spanish] El diseno bioclimatico aprovecha los recursos naturales para el mejoramiento de la calidad de la construccion y vida del usuario. La combinacion de estos dos conceptos dentro del diseno arquitectonico, genera dualidad interesante de entender el medio ambiente fisico y psicologico de los usuarios y obtener una identificacion entre la realidad y los suenos. La psicologia ambiental se ha enfocado a investigar sobre la relacion de los efectos psicologicos del ambiente construido o arquitectonico sobre la conducta y la experiencia humana. La conducta humana trabaja bajo parametros establecidos durante el desarrollo del mismo, a traves de los habitos y estilos de vida desde su infancia van marcando esos parametros que despues sirven para lograr satisfactores en la busqueda de la calidad de vida de los seres humanos.

Fuel radial design using Path Relinking; Diseno radial de combustible usando Path Relinking

Energy Technology Data Exchange (ETDEWEB)

Campos S, Y. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

2007-07-01

The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

Design and construction of a new furnace combustion pilot in IIE; Diseno y construccion de un nuevo horno experimental de combustion en el IIE

Energy Technology Data Exchange (ETDEWEB)

Martinez Flores, Marco Antonio; Tamayo Flores, G. A. [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)]. E-mail: mamf@iie.org.mx; gatamayo@iie.org.mx

2010-11-15

To expand the scope of studies and technological developments in the Laboratory of Combustion of the Thermal Process Management a new experimental furnace with a capacity of 1.5 MWt was designed and built. Spray design studies with real operational evaluation (comparison of emission of Combustion products and thermal behavior) as well as information on the behavior of the NO{sub x} emissions will be produced. An automatic control system using a water cooling jacket provides the cooling of the furnace. The furnace design is modular, consisting of six sections. Currently three trials to its maximum load of operation have been performed. The results show that the global energy balance of the furnace matches the design. The flue gas residence time and temperature also match those at the exit of steam generators of big capacity. [Spanish] Para ampliar las posibilidades de estudios y desarrollos tecnologicos del laboratorio de combustion de la Gerencia de Procesos Termicos, se diseno y construyo un nuevo horno experimental con capacidad maxima de 1.5 MegaWatts termicos (MWt). Este horno permitira realizar estudios de diseno de atomizadores con evaluacion de operacion real (comparacion de emision de productos de combustion y comportamiento termico), asi como ampliar la informacion que se tiene en el comportamiento de la emision de los oxidos de nitrogeno (NO{sub x}) que se producen en el horno. El diseno del horno es modular, compuesto por seis secciones, las cuales se enfrian de forma independiente a traves de un sistema de control automatico, empleando una chaqueta de enfriamiento por la cual circula agua. Actualmente se han realizado tres encendidos hasta alcanzar su carga maxima de operacion, los resultados obtenidos muestran que el balance de energia global del horno experimental es practicamente el de diseno y cumple con las caracteristicas de tiempo de residencia y temperatura de gases a la salida del horno de Generadores de Vapor de gran capacidad.

ITER fuel cycle

International Nuclear Information System (INIS)

Leger, D.; Dinner, P.; Yoshida, H.

1991-01-01

Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

Thermodynamic methodology for the design of solar dryers operated with flat solar collectors; Metodologia termodinamica para el diseno de secadores operados con calentadores solares planos

Energy Technology Data Exchange (ETDEWEB)

Torres Reyes, Ernestina; Navarrete Gonzalez, Jose L; Ibarra Salazar, Beatriz A; Picon Nunez, Martin [Instituto de Investigaciones Cientificas, Universidad de Guanajuato, Guanjuato, Guanajuato (Mexico)

2000-07-01

In this paper a thermal performance analysis of solar drying process at operating varying conditions is presented. It is described semi-empirical models to thermal characterization of an experimental device. A simulator of thermal performance for operating varying conditions was developed as a part of the procedure of thermal design of solar dryers. On the other hand, it is described a simplified method to design solar collectors based on the determination of minimum entropy generation during the thermal conversion of the solar device by using the thermal analysis procedure established and the method derived of the second law of the Thermodynamics are finally presented. [Spanish] En este trabajo se presenta el analisis termico del comportamiento del sistema -colector solar camara de secado-. Se describen los modelos semi-empiricos con los que se caracterizo termicamente un secador solar experimental del tipo indirecto. Se presenta tambien un procedimiento de diseno de equipo de secado que toma en cuenta las condiciones variables de operacion que presentan los dispositivos solares. Por otro lado se describe un procedimiento simplificado de diseno, basado en un analisis derivado de la segunda ley de la Termodinamica. Esta metodologia se fundamenta en la minima generacion de entropia durante la conversion termica de la energia solar, utilizando colectores solares planos. Finalmente se presentan los resultados del diseno preliminar de equipo de secado utilizando los dos procedimientos mencionados.

Thermoelectric power plant conversion from fuel oil to coal with integration of a CO{sub 2} capture plant. Part 1; Conversion de una central termoelectrica de combustoleo a carbon con integracion de una planta de captura de CO{sub 2}. Parte 1

Energy Technology Data Exchange (ETDEWEB)

Huante Perez, Liborio; Rodriguez Martinez, J. Hugo; Arriola Medellin, Alejandro M. [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)

2010-11-15

The major features in the design of power plant conversion from fuel oil to coal, considering the new technical characteristics of fuel and his transportation to plant, equipment required to comply with Mexican environmental standards and additional requirements of auxiliaries are described. In addition, changes needed on power plant design were considered according to integrates CO{sub 2} capture plant alternatives. [Spanish] En este articulo se describen las caracteristicas principales en el diseno de la conversion de Centrales de combustoleo a carbon, considerando el nuevo combustible y su transporte hasta la central, los equipos requeridos para cumplir con las normas ambientales y los requerimientos adicionales de los servicios auxiliares. Adicionalmente, se lleva a cabo el analisis de los cambios requeridos en el diseno de la conversion considerando diferentes opciones para la integracion de una planta de captura de CO{sub 2}, la cual debera entregar este subproducto para su compresion y envio a su destino final para su almacenamiento.

Implementation of the design attendent by computers (CAD) for the location of structures of power transmission lines; Implementacion del diseno asistido por computadora para la localizacion de estructuras de lineas de transmision

Energy Technology Data Exchange (ETDEWEB)

Vega Ortiz, Miguel; Gutierrez Arriola, Gustavo [Instituto de Investigaciones Electricas, Temixco, Morelos (Mexico)

2000-07-01

In order that the tools of CAD (Design Attended by Computer) that are offered in the market are really useful, they must combine the criteria and experiences of the expert designers with the specifications and practices established in the electrical company. This includes, from the introduction to the information system of the available input data and its design criteria, to obtaining the required output information. In the present work the methodology developed by the Instituto de Investigaciones Electricas (IIE) in the design of power transmission lines that integrates the Comision Federal de Electricidad (CFE) requirements in the design of its transmission power lines is an advanced computer tool that results in obtaining better designs. Some of the most important aspects are the reduction of the used working time, the cost of the designed line, its reliability, the flexibility in the information handling and the quality of presentation. [Spanish] Para que las herramientas de diseno asistido por computadora que se ofrecen en el mercado sean realmente utiles deben conjuntar los criterios y experiencias de los disenadores expertos con las especificaciones y practicas establecidas en la empresa electrica. Esto incluye desde la introduccion al sistema de la informacion de datos de entrada de la que se dispone y de sus criterios de diseno hasta la obtencion de la informacion de salida que se requiere. En el presente trabajo se resume la metodologia desarrollada por el Instituto de Investigaciones Electricas (IIE) en el diseno de lineas de transmision, que integra los requerimientos de la Comision Federal de Electricidad (CFE) en el diseno de sus lineas de transmision en una herramienta de computo avanzada y que redunda en la obtencion de mejores disenos. Algunos de los aspectos mas importantes son la reduccion del tiempo de trabajo empleado, el costo de la linea disenada, su confiabilidad, la flexibilidad en el manejo de informacion y la calidad de presentacion.

Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

International Nuclear Information System (INIS)

Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

2012-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

Conceptual design report for the away from reactor spent fuel storage facility, Savannah River Plant

International Nuclear Information System (INIS)

1978-12-01

The Department of Energy (DOE) requested that Du Pont prepare a conceptual design and appraisal of cost for Federal budget planning for an away from reactor spent fuel storage facility that could be ready to store fuel by December 1982. This report describes the basis of the appraisal of cost in the amount of $270,000,000 for all facilities. The proposed action is to provide a facility at the Savannah River Plant. The facility will have an initial storage capacity of 5000 metric tons of spent fuel and will be capable of receiving 1000 metric tons per year. The spent fuel will be stored in water-filled concrete basins that are lined with stainless steel. The modular construction of the facility will allow future expansion of the storage basins and auxiliary services in a cost-effective manner. The facility will be designed to receive, handle, decontaminate and reship spent fuel casks; to remove irradiated fuel from casks; to place the fuel in a storage basin; and to cool and control the quality of the water. The facility will also be designed to remove spent fuel from storage basins, load the spent fuel into shipping casks, decontaminated loaded casks and ship spent fuel. The facility requires a license by the Nuclear Regulatory Commission (NRC). Features of the design, construction and operations that may affect the health and safety of the workforce and the public will conform with NRC requirements. The facility would be ready to store fuel by January 1983, based on normal Du Pont design and construction practices for DOE. The schedule does not include the effect of licensing by the NRC. To maintain this option, preparation of the documents and investigation of a site at the Savannah River Plant, as required for licensing, were started in FY '78

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.

Energy saving in office buildings through cheap bioclimatic design measures; Ahorro de energia en edificios de oficinas a partir de medidas de diseno bioclimatico de bajo costo

Energy Technology Data Exchange (ETDEWEB)

Garcia Chavez, Jose Roberto [Universidad Autonoma Metropolitana, Unidad Azcapotzalco (Mexico)

2007-01-15

This article shows the study performed in a building of Campus Uni dad Azcapotzalco in order to evaluate some bioclimatic design strategies. And to reach the higrothermic comfort and the appropriate lighting to favor employees to increase their performance in work activities. There are some strategies in order to solve this problem: the skylight area reduction, existing lights change, ventilation increase, fixed partition door implementation, etc. The obtained outcomes make possible the beginning of bioclimatic design use to save energy in other buildings, because sometimes edifications possess a design that cause the energy waste. [Spanish] En este articulo se muestra el estudio que se hizo en un edificio en el Campus Unidad Azcapotzalco, el fin de esta labor es evaluar ciertas estrategias de diseno bioclimatico para que las personas que operan en este edificio puedan tener confort higrotermico y una iluminacion pertinente, que den como resultado un mejor desempeno en las tareas que realizan. Algunas de las estrategias para darle solucion a esta problematica son: disminuir del area de lucernarios, cambiar las luminarias existentes, aumentar la ventilacion, implementar canceleria, etc. Con los resultados obtenidos nace la esperanza de que se utilicen estrategias de diseno bioclimatico para ahorrar energia en otros edificios, ya que muchas veces las edificaciones poseen un diseno que unicamente provoca el uso imprudente de la energia.

La falta de financiacion publica hipoteca el futuro del acelerador de protones valenciano. La Universidad de Valencia tiene avanzado el diseno de este dispositivo de alta tecnologia

CERN Multimedia

Jatvia, J M

2002-01-01

"La Universidad de Valencia tiene avanzado el diseno de un acelerador lineal de protones, cuya culminacion depende del compromiso del Consell en sufragar la plantilla que necesitaria el centro para estar operativo" (1 page).

Substations grounding systems design; Diseno de sistemas de conexion a tierra en subestaciones

Energy Technology Data Exchange (ETDEWEB)

Velazquez Sanchez, Raul [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1987-12-31

The grounding systems have been the object of a great number of analytical and experimental studies in the past 15 years, that have given rise to research reports on the several aspects involved in this technique. In this work are briefly reviewed various aspects that is necessary to take into account for its design, analysis and put in operation. [Espanol] Los sistemas de conexion a tierra han sido objeto de gran cantidad de estudios analiticos y experimentales en los pasados quince anos, que han dado origen a informes de investigacion sobre los diversos aspectos que involucra esta tecnica. En este trabajo se revisan brevemente diversos aspectos que es necesario considerar para su diseno, analisis y puesta en operacion.

Substations grounding systems design; Diseno de sistemas de conexion a tierra en subestaciones

Energy Technology Data Exchange (ETDEWEB)

Velazquez Sanchez, Raul [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1986-12-31

The grounding systems have been the object of a great number of analytical and experimental studies in the past 15 years, that have given rise to research reports on the several aspects involved in this technique. In this work are briefly reviewed various aspects that is necessary to take into account for its design, analysis and put in operation. [Espanol] Los sistemas de conexion a tierra han sido objeto de gran cantidad de estudios analiticos y experimentales en los pasados quince anos, que han dado origen a informes de investigacion sobre los diversos aspectos que involucra esta tecnica. En este trabajo se revisan brevemente diversos aspectos que es necesario considerar para su diseno, analisis y puesta en operacion.

DEMONSTRATION OF FUEL CELLS TO RECOVER ENERGY FROM ANAEROBIC DIGESTER GAS - PHASE I. CONCEPTUAL DESIGN, PRELIMINARY COST, AND EVALUATION STUDY

Science.gov (United States)

The report discusses Phase I (a conceptual design, preliminary cost, and evaluation study) of a program to demonstrate the recovery of energy from waste methane produced by anaerobic digestion of waste water treatment sludge. The fuel cell is being used for this application becau...

Conceptual core designs for a 1200 MWe sodium cooled fast reactor

International Nuclear Information System (INIS)

Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

2008-01-01

The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

Energy savings in buildings through the architectonic design; Ahorro de energa en edificaciones a traves de diseno arquitectonico

Energy Technology Data Exchange (ETDEWEB)

Figueroa Castrejon, A. [Universidad Autnoma Metropolitana-Azcapotzalco, Division de Ciencias y Artes para el Diseno, Laboratorio de Arquitectura Bioclimatica (Mexico)

1997-12-31

This paper points out the problems the existing buildings have in its architectonic design and alternatives presented to improve their habitability and energy consumption, making reference to national cases and to the social-economic environment of our country. [Espanol] El presente trabajo senala los problemas que tienen las edificaciones actuales en su diseno arquitectonico y las alternativas que se presentan para mejorar su habitabilidad y consumo energetico, haciendo referencia a los casos nacionales y al entorno socioeconmico de nuestro pas.

Energy savings in buildings through the architectonic design; Ahorro de energa en edificaciones a traves de diseno arquitectonico

Energy Technology Data Exchange (ETDEWEB)

Figueroa Castrejon, A [Universidad Autnoma Metropolitana-Azcapotzalco, Division de Ciencias y Artes para el Diseno, Laboratorio de Arquitectura Bioclimatica (Mexico)

1998-12-31

This paper points out the problems the existing buildings have in its architectonic design and alternatives presented to improve their habitability and energy consumption, making reference to national cases and to the social-economic environment of our country. [Espanol] El presente trabajo senala los problemas que tienen las edificaciones actuales en su diseno arquitectonico y las alternativas que se presentan para mejorar su habitabilidad y consumo energetico, haciendo referencia a los casos nacionales y al entorno socioeconmico de nuestro pas.

Conceptual aspects of the safety evaluation of a project of complementary spent nuclear fuel dry storage unit

Energy Technology Data Exchange (ETDEWEB)

Freitas, Rafaela da S. A.; Fontes, Gladson S., E-mail: rafaaelaandrade@hotmail.com, E-mail: gsfontes@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Saldanha, Pedro L. C., E-mail: saldanha@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

Based on the number of cycles and the amount of new fuel elements exchanged in the reactor cores at each cycle, the forecast for the exhaustion of the spent nuclear fuel pools of the Brazil plants has provision until 2021. As are still in the studies the availability of a long-term storage facility for spent fuel, the short-term solution will be the construction of the Complementary Storage Spent Nuclear Fuel Unit, it will build inside the site in Angra Plants. The dry cask is a method of storage in which the fuel elements of high-level radioactive waste are stored, such as spent nuclear fuel, which already cooled in the fuel pool for at least one year and up to ten years. The purpose of the present paper is to discuss a conceptual study of the safety analysis of a project of licensing of a Dry Storage Unit (DSU) with the objective of verifying the application of national and international criteria, requirements and standards. The safety analysis will make on the principles adopted by the US Nuclear USNRC and the standards adopted at CNEN for dry storage. The concept of installation, seismic, geological and other analysis will be approached for approval of the site to be installed at DSU, the approved permit for the construction and finally the external and internal events that may occur being incidents and / or accidents and which are The necessary mitigations if something occurs within a period of time. (author)

Conceptual aspects of the safety evaluation of a project of complementary spent nuclear fuel dry storage unit

International Nuclear Information System (INIS)

Freitas, Rafaela da S. A.; Fontes, Gladson S.; Saldanha, Pedro L. C.

2017-01-01

Based on the number of cycles and the amount of new fuel elements exchanged in the reactor cores at each cycle, the forecast for the exhaustion of the spent nuclear fuel pools of the Brazil plants has provision until 2021. As are still in the studies the availability of a long-term storage facility for spent fuel, the short-term solution will be the construction of the Complementary Storage Spent Nuclear Fuel Unit, it will build inside the site in Angra Plants. The dry cask is a method of storage in which the fuel elements of high-level radioactive waste are stored, such as spent nuclear fuel, which already cooled in the fuel pool for at least one year and up to ten years. The purpose of the present paper is to discuss a conceptual study of the safety analysis of a project of licensing of a Dry Storage Unit (DSU) with the objective of verifying the application of national and international criteria, requirements and standards. The safety analysis will make on the principles adopted by the US Nuclear USNRC and the standards adopted at CNEN for dry storage. The concept of installation, seismic, geological and other analysis will be approached for approval of the site to be installed at DSU, the approved permit for the construction and finally the external and internal events that may occur being incidents and / or accidents and which are The necessary mitigations if something occurs within a period of time. (author)

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sease, J.D.; Primm, R.T. III; Miller, J.H.

2007-01-01

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Sease, J.D.; Primm, R.T. III; Miller, J.H.

2007-09-30

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are

Conceptual Biorefinery Design and Research Targeted for 2022: Hydrothermal Liquefacation Processing of Wet Waste to Fuels

Energy Technology Data Exchange (ETDEWEB)

Snowden-Swan, Lesley J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhu, Yunhua [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bearden, Mark D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Seiple, Timothy E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Susanne B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schmidt, Andrew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billing, Justin M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hallen, Richard T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hart, Todd R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Jian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Albrecht, Karl O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fox, Samuel P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maupin, Gary D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elliott, Douglas C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-12-28

The Department of Energy Bioenergy Technologies Office (BETO) invests in research and development of new pathways for commercially viable conversion of biomass into drop-in ready transportation fuels, fuel blendstocks and products. The primary emphasis has been on terrestrial and algae feedstocks, but more recently BETO has begun to explore the potential of wet wastes for biofuel production, with focus on wastewater residuals, manure, food waste, and fats, oils and grease. A recent resource analysis estimates that 77 million dry tons per year of these wastes are generated annually, 65% of which are underutilized for any beneficial purpose. Approximately 14 million dry tons of the total resource is wastewater residuals (sludge and biosolids) generated at the nation’s wastewater treatment plants (WWTPs). Conversion of this resource into transportation fuels could significantly contribute to the creation of a new domestic bioenergy and bioproduct industry, while providing an economically and environmentally sustainable alternative for current waste disposal practices. Hydrothermal liquefaction (HTL) is a process that uses hot, pressurized water in the condensed phase to convert biomass to a thermally stable oil product, also known as “biocrude”, which can then be thermo-catalytically upgraded to hydrocarbon fuel blendstocks. HTL is conceptually simple, has a high carbon efficiency, and can be applied to a wide range of wet feedstocks at similar processing conditions. The purpose of this report is to document the conceptual design, economics and supporting data for a sludge-to-fuel pathway via HTL and biocrude upgrading. The configuration includes a HTL plant that is co-located with a WWTP and a larger scale biocrude upgrading plant for production of hydrocarbon fuel blendstocks. Experimental data from bench scale testing of a 1:1 mixture of primary:secondary sludges are used to establish the economic and technical assumptions for the analysis. The design

NWTS conceptual reference repository description (CRRD). Volume 1. Summary

International Nuclear Information System (INIS)

1981-05-01

This report is a Conceptual Reference Repository Description (CRRD) for the terminal storage of spent, unreprocessed fuel assemblies and low-level transuranic waste. It is based on the following three previously prepared conceptual design reports: National Waste Terminal Storage in a Bedded Salt Formation for Spent Unreprocessed Fuel (NWTS-R2), prepared by Kaiser Engineers; National Waste Terminal Storage Repository Number 1 (NWTS-R1), prepared by Stearns-Roger; Spent Fuel Receiving and Packaging Facility Conceptual Design, prepared by Rockwell Hanford Operations and Kaiser Engineers. The CRRD is not itself a conceptual design report, but combines the key features of the above three reports into a description of a reference repository. This description is to be used as a basis for the preparation of a preliminary information report for the repositories. The CRRD discusses from a conceptual engineering standpoint the structures, systems, equipment, and operations necessary to (1) receive unreprocessed spent fuel assemblies via standard shipping casks, (2) offload these assemblies, (3) place the assemblies in canisters, (4) transport the canisters to underground storage locations in the salt dome, and (5) place these canisters in terminal storage. The CRRD also elaborates on the concepts for the retrieval and recovery of the spent fuel after burial; describes the development of the shafts and the underground areas,as well as the supporting operational utility and administrative features of the repository; and discusses the impacts on the reference repository of (1) changes in the waste package concept, (2) dry receiving and dry temporary storage of waste, (3) terminal storage of reprocessing waste, and (4) alternative underground development schemes

Some safety studies for conceptual fusion--fission hybrid reactors. Final report

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.

1978-07-01

The objective of this study was to make a preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors. The study and subsequent analysis was largely based upon reference to one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The blanket is a fast-spectrum, uranium carbide, helium cooled, subcritical reactor, optimized for the production of fissile fuel. An attempt was made to generalize the results wherever possible

Conceptual design report of the SMART fuel rod

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

The SMART fuel rod is based on 17 x 17 KOFA(Korea Fuel Assembly) fuel rod of the 950MWe pressurize water reactor. The fuel stack length of the KOFA is 3658mm, otherwise SMART fuel rod stack length is 2000mm. The fuel rod contains UO{sub 2} pellets with the enrichment of 4.95%. All the fuel in core will be replaced every 35 months. The average LHGR of the fuel rod is 120 W/cm, commercial PWR is 178 W/cm, SMART LHGR is lower about 31% than commercial PWR. The core inlet and outlet temperature of coolant are respectively 270 deg C and 310 deg C, commercial PWR are respectively 291.6 deg C and 326.8 deg C, SMART inlet and outlet temperature is lower averaged 19.2 deg C than commercial PWR. The coolant use mixed soluble ammonia in high purity water and boron is not in. The general performance of the fuel rod UO{sub 2} pellet has been already verified through the sufficient burnup (60,000 MWd/MTU-rod avg.) experience as the rods of same design in commercial PWR's. But cladding corrosion is required the further verification. (author). 13 refs., 3 figs., 8 tabs.

Used fuel packing plant for CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

2015-07-01

Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

Nuclear fuel

International Nuclear Information System (INIS)

D Hondt, P.

1998-01-01

The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

National waste terminal storage conceptual reference repository description

International Nuclear Information System (INIS)

Odgers, I.L.; Collings, J.L.

1981-01-01

The conceptual reference repository description (CRRD) discusses, from a conceptual engineering standpoint, the structures, systems, equipment, and operations necessary to: (1) receive unreprocessed spent fuel assemblies in standard casks; (2) unload these assemblies; (3) place them in canisters; (4) transport the canisters to underground storage locations in the salt dome; and (5) place the canisters in terminal storage. The CRRD also elaborates on the concepts for retrieval and recovery of the spent fuel after burial and describes the development of the shafts and the underground areas, as well as the supporting operational utility and administrative features of the repository

Gamma scanner conceptual design report

International Nuclear Information System (INIS)

Swinth, K.L.

1979-11-01

The Fuels and Materials Examination Facility (FMEF) will include several stations for the nondestructive examination of irradiated fuels. One of these stations will be the gamma scanner which will be employed to detect gamma radiation from the irradiated fuel pins. The conceptual design of the gamma scan station is described. The gamma scanner will use a Standard Exam Stage (SES) as a positioner and transport mechanism for the fuel pins which it will obtain from a magazine. A pin guide mechanism mounted on the face of the collimator will assure that the fuel pins remain in front of the collimator during scanning. The collimator has remotely adjustable tungsten slits and can be manually rotated to align the slit at various angles. A shielded detector cart located in the operating corridor holds an intrinsic germanium detector and associated sodium-iodide anticoincidence detector. The electronics associated with the counting system consist of standard NIM modules to process the detector signals and a stand-alone multichannel analyzer (MCA) for counting data accumulation. Data from the MCA are bussed to the station computer for analysis and storage on magnetic tape. The station computer controls the collimator, the MCA, a source positioner and the SES through CAMAC-based interface hardware. Most of the electronic hardware is commercially available but some interfaces will require development. Conceptual drawings are included for mechanical hardware that must be designed and fabricated

Fuel cell design using a new heuristic method; Diseno de celdas de combustible mediante un nuevo metodo heuristico

Energy Technology Data Exchange (ETDEWEB)

Perusquia, R.; Montes T, J. L.; Ortiz S, J. J.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

In this paper a new method for the pre-design of a typical fuel cell with a structural array of 10 x 10 fuel elements for a BWR is presented. The method is based on principles of maximum dispersion and minimum peaks of local power within the array of fuel elements. The pre-design of the fuel cells is made by simulation in two dimensions (2-D) through the cells physics code CASMO-4. For this purpose of pre-design the search process is guided by an objective function which is a combination of the main neutronic parameters of the fuel cell. The results show that the method is a promising tool that could be used for the design of fuel cells for use in a nuclear plant BWR. (Author)

Analysis of the TREAT LEU Conceptual Design

Energy Technology Data Exchange (ETDEWEB)

Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-01

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO₂-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.

Conceptual design study of LMFBR core with carbide fuel

International Nuclear Information System (INIS)

Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

1987-01-01

Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

Prevention of criticality accidents. Fuel elements storage; Prevencion de accidentes de criticidad. Almacenamiento de elementos combustibles

Energy Technology Data Exchange (ETDEWEB)

Canavese, S I; Capadona, N M

1991-12-31

Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author). [Espanol] Partiendo de la necesidad de almacenar elementos combustibles tipo placa MTR (Materials Testing Reactors), producidos con uranio enriquecido al 20% en U235 para reactores de investigacion, se requiere el diseno de un deposito para tal fin que brinde esencialmente un alto grado de seguridad intrinseca y que no ofrezca complicaciones en cuanto a su construccion. (Autor).

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

International Nuclear Information System (INIS)

Monado, F.; Permana, S.

2013-01-01

Full-text: A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8 % HM. From the neutronic point of view, this design is in compliance with good performance. (author)

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

International Nuclear Information System (INIS)

Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

2014-01-01

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance

Intelligent buildings vs. bioclimatic design; Edificios inteligentes vs. diseno bioclimatico

Energy Technology Data Exchange (ETDEWEB)

Aguayo Gonzalez, Ricardo [Tecnologico de Monterrey (Mexico)

2006-10-15

Present the form of intelligent buildings designing is the article purpose. Those kinds of edifications take advantage of climatic conditions which allow the users comfort and the efficient electric power use, avoiding the polluting agents. It also shows the four next following stages to design an intelligent building to know: the dry weather and relative dampness schedule variations during a year in the building location; the predominant winds direction, intensity and schedule frequency; the cloudiness, rain, etc and how to use the Givoni diagram to obtain the natural air-conditioning strategies and reach the thermal comfort. [Spanish] El proposito de este articulo es presentar la forma de disenar edificios Inteligentes, los cuales aprovechan las condiciones climaticas que permiten el confort de los usuarios y el uso eficiente de la energia electrica, evitando asi la emision de agentes contaminantes. Tambien menciona los siguientes cuatro pasos para el diseno de un edificio inteligente: conocer las variaciones horarias de temperatura seca y humedad relativa durante un ano en el lugar donde estara el edificio, saber la direccion, intensidad y frecuencia horaria de los vientos dominantes, tener conocimiento de la nubosidad, lluvia, etc. y utilizar el diagrama de Givoni para obtener las estrategias de climatizacion natural para obtener el confort termico.

Catalyzed deuterium fueled reversed-field pinch reactor assessment

International Nuclear Information System (INIS)

Dobrott, D.

1985-01-01

This study is part of a Department of Energy supported alternate fusion fuels program at Science Applications International Corporation. The purpose of this portion of the study is to perform an assessment of a conceptual compact reversed-field pinch reactor (CRFPR) that is fueled by the catalyzed-deuterium (Cat-d) fuel cycle with respect to physics, technology, safety, and cost. The Cat-d CRFPR is compared to a d-t fueled fusion reactor with respect to several issues in this study. The comparison includes cost, reactor performance, and technology requirements for a Cat-d fueled CRFPR and a comparable cost-optimized d-t fueled conceptual design developed by LANL

Voltage controller design for air conditioning; Diseno de controlador de voltaje para aire acondicionado

Energy Technology Data Exchange (ETDEWEB)

Garcia Andrade, R; Lopez Villalobos, J.J; Valderrama Chairez, J; Ramirez, R.L. [Instituto Tecnologico de Nuevo Leon, Guadalupe, Nuevo Leon (Mexico)]. E-mails: roxana_garciaandrade@yahoo.com; xe2n@yahoo.com.mx; jose.valderrama@ieee.org

2013-03-15

This paper discusses the design of a voltage controller for an air conditioning system in order to generate additional power in activation or startup of the system, for which as a first stage is presented the modeling power generation of electric current through alternative means, such as solar energy. The results of this study will be the basis for development of the physical prototype of this system controller. [Spanish] El presente trabajo trata sobre el diseno de un controlador de voltaje para un sistema de aire acondicionado con el fin de generar energia adicional en la activacion o arranque de dicho sistema, para lo cual como primer fase se presenta el modelado de la generacion de corriente electrica mediante medios alternos, como lo es la energia solar. Los resultados de este trabajo seran la base para desarrollo del prototipo fisico de este sistema controlador.

Objectives of WPA. Conceptual and experimental requirements for the development of the MAN

International Nuclear Information System (INIS)

Bruno, J.

2003-01-01

In the framework of the SFS (Spent Fuel Stability) European project, a 3-day workshop has been organised to share ideas and get a common understanding in the issue of modelling the behaviour of the spent fuel under repository conditions. The ultimate goal is to set the experimental and theoretical basis for the development of the conceptual model for the fuel matrix alteration. The workshop has been organised with plenty of time to discuss the key issues both formally (presentations) and informally (meetings and discussions). Mos importantly, the presentations have been prepared in order to feed the construction and development of the conceptual model for the spent nuclear fuel matrix alteration, which is the ultimate goal of Work Package 4 of the SFS EU project. We do not want a standard Spent Fuel Workshop presentation, we really need your better experimental and theoretical ideas and realisation to accomplish the objective of this workshop. At the onset of the Workshop I will put for war our current thoughts on the development of a common conceptual model. Which are the main boundary conditions, what are the needed parameters and functionalities. Where are the main uncertainties, how can we narrow down these uncertainties. I expect the contributions from all the participants in order to feed back on these requirements and on the advancement towards a common conceptual model for SFS matrix alteration. (Author)

Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

International Nuclear Information System (INIS)

Mohd Fazli Zakaria; Tonny Lanyau; Ahmad Nabil Ab Rahim

2010-01-01

Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

Cost-Benefit of the energy saving in the bioclimatic designs; Costo-Beneficio del ahorro de energia en los disenos bioclimaticos

Energy Technology Data Exchange (ETDEWEB)

Lopez P, J. Manuel A. [Doble Dos Soluciones de Negocios, S.A. de C.V., Mexico, D.F. (Mexico); Morillon G, David; Rodriguez V, Luis [Instituto de Ingenieria de la UNAM, Mexico, D.F. (Mexico)

1999-07-01

Tools and criteria for the Cost-Benefit analysis from the energy saving in bioclimatic designs (BD) are presented, for which, a reference project is established and the costs and benefits of energy saving in BD with base in that project are evaluated. A case study is presented taking as reference the traditional design of the air conditioning of a building and with this base the estimation of cost-benefit of the same building is made, but with passive systems. The tools used are those that allow to consider related resources such as time and money; in that sense, are used criteria such as: Present value (PV), Equivalent Annual Cost (EAC) and the Cost of the Conserved Energy (CCE). The costs related to the construction, maintenance and operation of the design are taken into account. The differences between the reference design and the BD, established from the application of these criteria, allow to evaluate the economic margin of the BD as far as operation and maintenance is concerned. Finally the CCE of the bioclimatic measures is evaluated as an initiative of energy saving and the present value is calculated of the energy saving in the entire useful life of a bioclimatic design. [Spanish] Se presentan herramientas y criterios para el analisis Costo-Beneficio del ahorro de energia en disenos bioclimaticos (DB), para lo cual, se establece un proyecto de referencia y se evaluan los costos y beneficios del ahorro de energia en DB con base en ese proyecto. Se presenta el estudio de un caso tomando como referencia el dise tradicional de la climatizacion de un edificio y con base en este, se realiza la estimacion de costos y beneficios del mismo edificio, pero con sistemas pasivos. Las herramientas usadas, son aquellas que permiten considerar recursos relacionados como son tiempo y dinero; en ese sentido, se usan criterios como: Valor Presente (VP), Costo Anual Equivalente (CAE) y el Costo de la Energia Conservada (CEC). Se toman en consideracion los costos relacionados

Solar fuels and chemicals system design study (ammonia/nitric acid production process). Volume 2. Conceptual design. Final report

Energy Technology Data Exchange (ETDEWEB)

1986-06-01

As part of the Solar Central Receiver Fuels and Chemicals Program, Foster Wheeler Solar Development Corporation (FWSDC), under contract to Sandia National Laboratories-Livermore (SNLL), developed a conceptual design of a facility to produce ammonia and nitric acid using solar energy as the principal external source of process heat. In the selected process, ammonia is produced in an endothermic reaction within a steam methane (natural gas) reformer. The heat of reaction is provided by molten carbonate salt heated by both a solar central receiver and an exothermic ammonia-fired heater. After absorption by water, the product of the latter reaction is nitric acid.

MINIMARS conceptual design: Final report

International Nuclear Information System (INIS)

Lee, J.D.

1986-09-01

This volume contains the following sections: (1) fueling systems; (2) blanket; (3) alternative blanket concepts; (4) halo scraper/direct converter system study and final conceptual design; (5) heat-transport and power-conversion systems; (6) tritium systems; (7) minimars air detritiation system; (8) appropriate radiological safety design criteria; and (9) cost estimate

Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

International Nuclear Information System (INIS)

Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

2010-01-01

The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

Hydraulic modelling of the CARA Fuel element; Desarrollo hidraulico del combustible CARA

Energy Technology Data Exchange (ETDEWEB)

Brasnarof, Daniel O; Juanico, Luis [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Disenios Avanzados y Evaluacion Economica; Giorgi, M [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales; Ghiselli, Alberto M; Zampach, Ruben; Fiori, Jose M; Yedros, Pablo A [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Ensayos no Destructivos

2004-07-01

The CARA fuel element is been developing by the National Atomic Energy Commission for both Argentinean PHWRs. In order to keep the hydraulic restriction in their fuel channels, one of CARA's goals is to keep its similarity with both present fuel elements. In this paper is presented pressure drop test performed at a low-pressure facility (Reynolds numbers between 5x10{sup 4} and 1,5x10{sup 5}) and rational base models for their spacer grid and rod assembly. Using these models, we could estimate the CARA hydraulic performance in reactor conditions that have shown to be satisfactory. (author) [Spanish] Con el objeto de validar la similitud hidraulica del elemento combustible CARA con los actuales combustibles de Atucha y Embalse, se realizaron ensayos de perdida de carga en el circuito CBP del CAC con un nuevo diseno de separador de mejor desempeno hidraulico. Se presenta aqui el analisis de los mismos, de los cuales se validaron modelos de base racional para estimar las restricciones hidraulicas de los distintos componentes estructurales (separadores, grillas y barras combustibles) en funcion del flujo refrigerante. Se estimo asi la caida de presion del CARA dentro del canal combustible Embalse en condiciones nominales de reactor, siendo la misma similar al del combustible actual de 37 barras. (autor)

Structural Analysis in a Conceptual Design Framework

Science.gov (United States)

Padula, Sharon L.; Robinson, Jay H.; Eldred, Lloyd B.

2012-01-01

Supersonic aircraft designers must shape the outer mold line of the aircraft to improve multiple objectives, such as mission performance, cruise efficiency, and sonic-boom signatures. Conceptual designers have demonstrated an ability to assess these objectives for a large number of candidate designs. Other critical objectives and constraints, such as weight, fuel volume, aeroelastic effects, and structural soundness, are more difficult to address during the conceptual design process. The present research adds both static structural analysis and sizing to an existing conceptual design framework. The ultimate goal is to include structural analysis in the multidisciplinary optimization of a supersonic aircraft. Progress towards that goal is discussed and demonstrated.

General features of conceptual design for the pilot plant to manufacture fuel rods from mixed oxides

International Nuclear Information System (INIS)

Quesada, C.A.; Adelfang, P.; Esteban, A.; Aparicio, G.; Friedenthal, M.; Orlando, O.S.

1987-01-01

This paper conceptually describes: 1) the processes in the manufacturing lines; 2) the distribution of quality controls and glove boxes in manufacturing lines; 3) the Control and Radiological Safety Room; 4) the Dressing Room; 5) the requirements of the ventilation system. The plant will be located in the first floor of the Radiochemical Processes Laboratory building, occupying a surface of 600 m 2 . The necessary equipment for the following manufacturing lines will be provided: a) conversion from Pu(NO3)4 to PuO 2 (through Pu(III)oxalate); b) manufacture of homogeneous of mixed oxides of U and Pu; c) manufacture of (U,Pu)O 2 pellets; d) manufacture of fuel rods of mixed uranium and plutonium oxides. (Author)

GFR fuel and core pre-conceptual design studies

International Nuclear Information System (INIS)

Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M.; Malo, J.Y.; Garnier, J.C.; Bertrand, F.; Bosq, J.C.

2007-01-01

The revision of the GFR core design - plate type - has been undertaken since previous core presented at Global'05. The self-breeding searched for has been achieved with an optimized design ('12/06 E'). The higher core pressure drop was a matter of concern. First of all, the core coolability in natural circulation for pressurized conditions has been studied and preliminary plant transient calculations have been performed. The design and safety criteria are met but no more margin remains. The project is also addressing the feasibility and the design of the fuel S/A. The hexagonal shape together with the principle of closed S/A (wrapper tube) is kept. Ceramic plate type fuel element combines a high enough core power density (minimization of the Pu inventory) and plutonium and minor actinides recycling capabilities. Innovative for many aspects, the fuel element is central to the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is considered as the other fuel element of interest. This combination of fuel/core feasibility and performance analysis, safety dispositions and performances analysis will compose the 'GFR preliminary feasibility' which is a project milestone at the end of the year 2007. (authors)

Spent fuel storage and isolation

International Nuclear Information System (INIS)

Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

1979-02-01

The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

A study on the safety of spent fuel management. A scenario study on spent fuel management

Energy Technology Data Exchange (ETDEWEB)

Chun, Kwan Sik; Park, Hyun Soo; Ahn, Jin Soo; Hwang, Joo Ho; Choi, Jong Won; Kim, Yeon Soo; Park, Ju Hwan; Chung, Choong Hwan [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

1992-03-01

In order to produce data applicable for the long-term policy making of spent fuel management and to suggest a basic scenario suitable to domestic situation, the pre-conceptual design of reference disposal facilities for the spent fuel and the vitrified high level radioactive waste from its reprocessing, has been performed. From the results of the pre-conceptual study, further research and development areas to accumulate the disposal technology are suggested. In addition, the physico-chemical properties and functional characteristics of domestic bentonite are analyzed to assess its applicability as a buffer material which would play a major role for the safe disposal of highly active waste including spent fuels. (Author).

Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

Energy Technology Data Exchange (ETDEWEB)

Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Baek, J. S [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.

2014-04-30

It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

Diseno de una Actividad de Aprendizaje Basada en la Argumentacion Dialogica en un curso Virtual de Biotecnologia y su Incidencia en el Desarrollo de Competencias Cientificas

Science.gov (United States)

Ortiz Benavides, Fedra Lorena

El proposito de la investigacion fue evaluar la efectividad de una actividad de aprendizaje basado en la argumentacion dialogica en linea y su incidencia en el desarrollo de competencias cientificas. Se fundamenta en la teoria del aprendizaje socio cultural de Vigotsky (1984), los principios del diseno instruccional de la cognicion situada por Hung y Der-Thang (2001) y como estrategia se aplico la argumentacion dialogica utilizando el Modelo Argumentativo de Toulmin MAT (1984). El diseno experimental comparo dos grupos de estudiantes A y B en el curso virtual de Biotecnologia. El grupo A (experimental) desarrollo la discusion a partir de la estrategia disenada para este estudio y el grupo B (control) realizo la discusion desde las actividades tradicionales. El desarrollo de la competencia argumentativa se valoro con el instrumento de evaluacion para argumentacion dialogica en linea propuesta por Clark y Sampson (2008). La evaluacion de las competencias cientificas se realizo a partir de una postprueba. Los datos fueron analizados con pruebas estadisticas no parametricas. Los resultados de la investigacion, indicaron diferencias significativas en el nivel de la competencia argumental en el grupo experimental en comparacion al grupo control. Igualmente se demostro que existe una relacion positiva entre el nivel de desarrollo de la competencia argumentativa y el nivel de desarrollo de las competencias cientificas.

Integral Monitored Retrievable Storage (MRS) Facility conceptual basis for design

International Nuclear Information System (INIS)

1985-10-01

The purpose of the Conceptual Basis for Design is to provide a control document that establishes the basis for executing the conceptual design of the Integral Monitored Retrievable Storage (MRS) Facility. This conceptual design shall provide the basis for preparation of a proposal to Congress by the Department of Energy (DOE) for construction of one or more MRS Facilities for storage of spent nuclear fuel, high-level radioactive waste, and transuranic (TRU) waste. 4 figs., 25 tabs

Axial design of fuel for BWRs using neural networks; Diseno axial de combustible para BWRs usando redes neuronales

Energy Technology Data Exchange (ETDEWEB)

Ortiz, J.J.; Castillo, A.; Montes, J.L.; Perusquia, R. [ININ, Carretera Mexico-Toluca s/n, 52750 La Marquesa, Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jjortiz@nuclear.inin.mx

2007-07-01

In this work a new system of axial optimization of fuel is presented based on a recurrent multi state neural net called RENODC. They are described with detail the main characteristics of this type of neural net (architecture, energy function and actualization of neural states) and like was adapted to the assemble design of nuclear fuel. The fuel design is proven by means of a fuel recharge and pre determined control rod patterns. By this way a good axial fuel design one has, when the thermal limits are fulfilled along the cycle, the reactor stays critic and at least the wanted longitude of the cycle is reached; also the margin of in cold turned off is verified. The assemble of fuel created with RENODC it is substituted by a recharge assemble and it is sought to verify that the energy requirements and aspects of safety are completed. The used cycle corresponds to a balance cycle of 18 months that it can be applied to the Laguna Verde Nucleo electric Central. The tests demonstrate the effectiveness of the system to reach satisfactory results in times of CPU of around 4 hours. This way, it could be proven that the design proposed with a lightly superior enrichment to that of the substituted design, fulfills the energy requirements. In later stages of this project this system will be coupled to the other optimization modules that are already had. (Author)

Conceptual design of cost-effective and environmentally-friendly configurations for fuel ethanol production from sugarcane by knowledge-based process synthesis.

Science.gov (United States)

Sánchez, Óscar J; Cardona, Carlos A

2012-01-01

In this work, the hierarchical decomposition methodology was used to conceptually design the production of fuel ethanol from sugarcane. The decomposition of the process into six levels of analysis was carried out. Several options of technological configurations were assessed in each level considering economic and environmental criteria. The most promising alternatives were chosen rejecting the ones with a least favorable performance. Aspen Plus was employed for simulation of each one of the technological configurations studied. Aspen Icarus was used for economic evaluation of each configuration, and WAR algorithm was utilized for calculation of the environmental criterion. The results obtained showed that the most suitable synthesized flowsheet involves the continuous cultivation of Zymomonas mobilis with cane juice as substrate and including cell recycling and the ethanol dehydration by molecular sieves. The proposed strategy demonstrated to be a powerful tool for conceptual design of biotechnological processes considering both techno-economic and environmental indicators. Copyright © 2011 Elsevier Ltd. All rights reserved.

Conceptual development of a test facility for spent fuel management

Energy Technology Data Exchange (ETDEWEB)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

Conceptual development of a test facility for spent fuel management

International Nuclear Information System (INIS)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

1997-01-01

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

Preliminary conceptual study of engineering-scale pyroprocess demonstration facility

International Nuclear Information System (INIS)

Moon, Seong-In; Chong, Won-Myung; You, Gil-Sung; Ku, Jeong-Hoe; Kim, Ho-Dong

2013-01-01

Highlights: ► The conceptual design of a pyroprocess demonstration facility was performed. ► The design requirements for the pyroprocess hot cell and equipment were determined. ► The maintenance concept for the pyroprocess hot cell was presented. -- Abstract: The development of an effective management technology of spent fuel is important to enhance environmental friendliness, cost viability and proliferation resistance. In Korea, pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems. PRIDE (PyRoprocess Integrated inactive DEmonstration facility) has been developed from 2007 to 2012 in Korea as a cold test facility to support integrated pyroprocessing and an equipment demonstration, which is essential to verify the pyroprocess technology. As the next stage of PRIDE, the design requirements of an engineering-scale demonstration facility are being developed, and the preliminary conceptual design of the facility is being performed for the future. In this paper, the main design requirements for the engineering-scale pyroprocess demonstration facility were studied in the throughput of 10tHM a year. For the preliminary conceptual design of the facility, the design basis of the pyroprocess hot cell was suggested, and the main equipment, main process area, operation area, maintenance area, and so on were arranged in consideration of the effective operation of the hot cells. Also, the argon system was designed to provide and maintain a proper inert environment for the pyroprocess. The preliminary conceptual design data will be used to review the validity of the engineering-scale pyroprocess demonstration facility that enhances both safety and nonproliferation

On the safety of conceptual fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

1979-01-01

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

Pre-conceptual core design of SCWR with annular fuel rods

Energy Technology Data Exchange (ETDEWEB)

Zhao, Chuanqi [Key Laboratory of Thermo-Fluid Science and Engineering of MOE, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Cao, Liangzhi, E-mail: caolz@mail.xjtu.edu.cn [Key Laboratory of Thermo-Fluid Science and Engineering of MOE, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Wu, Hongchun; Zheng, Youqi [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China)

2014-02-15

Highlights: • Annular fuel with both internal and external cooling is used in supercritical light water reactor (SCWR). • The geometry of the annular fuel has been optimized to achieve better performance for the SCWR. • Based on the annular fuel assembly, an equilibrium core has been designed. • The results show that the equilibrium core has satisfied all the objectives and design criteria. - Abstract: The new design of supercritical light water reactor was proposed using annular fuel assemblies. Annular fuel consists of several concentric rings. Feed water flows through the center and outside of the fuel to give both internal and external cooling. Thanks to this feature, the fuel center temperature and the cladding temperature can be reduced and high power density can be achieved. The water flowing through the center also provides moderation, so there is no need for extra water rods in the assembly. The power distribution can be easily flattened by use of this design. The geometry of the annular fuel has been optimized to achieve better performance for the SCWR. There are 19 fuel pins in an assembly. Burnable poison is utilized to reduce the initial excess reactivity. The fuel reloading pattern and water flow scheme were optimized to achieve more uniform power distribution and lower cladding temperature. An equilibrium core has been designed and analyzed using three dimensional neutronics and thermal-hydraulics coupling calculations. The void reactivity, Doppler coefficient and cold shut down margin were calculated for safety consideration. The present results show that this concept is a promising design for the SCWR.

Accelerator-driven subcritical facility:Conceptual design development

Science.gov (United States)

Gohar, Yousry; Bolshinsky, Igor; Naberezhnev, Dmitry; Duo, Jose; Belch, Henry; Bailey, James

2006-06-01

A conceptual design development of an accelerator-driven subcritical facility has been carried out in the preparation of a joint activity with Kharkov Institute of Physics and Technology of Ukraine. The main functions of the facility are the medical isotope production and the support of the Ukraine nuclear industry. An electron accelerator is considered to drive the subcritical assembly. The neutron source intensity and spectrum have been studied. The energy deposition, spatial neutron generation, neutron utilization fraction, and target dimensions have been quantified to define the main target performance parameters, and to select the target material and beam parameters. Different target conceptual designs have been developed based the engineering requirements including heat transfer, thermal hydraulics, structure, and material issues. The subcritical assembly is designed to obtain the highest possible neutron flux level with a Keff of 0.98. Different fuel materials, uranium enrichments, and reflector materials are considered in the design process. The possibility of using low enrichment uranium without penalizing the facility performance is carefully evaluated. The mechanical design of the facility has been developed to maximize its utility and minimize the time for replacing the target and the fuel assemblies. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements, upgrades, and new missions. In addition, it has large design margins to accommodate different operating conditions and parameters. In this paper, the conceptual design and the design analyses of the facility will be presented.

Design of earth network in electric substations; Diseno de redes de tierra en subestaciones electricas

Energy Technology Data Exchange (ETDEWEB)

Raull-Martin, J. [Universidad Nacional Autonoma de Mexico, Mexico, D.F. (Mexico)

2000-09-01

The purpose of this paper is to present a summarized version of the different types of construction of grounding systems, as web as their materials and formulas, which are necessary and useful for the design and building of a grounding network for medium to high voltage facilities. A description is also made of how to determine the electrical resistivity of subsoil layers with different physical characteristics. Several illustrative examples are also solved in the paper. [Spanish] El proposito de este articulo es presentar de manera resumida los diferentes tipos de construccion de los sistemas de tierra, asi como sus materiales y formulas, los cuales son necesarios y utiles para el diseno y edificacion de una red de tierra para instalaciones de mediana y alta tension. Se describe tambien la obtencion de las diferentes resistividades que presentan los terrenos con diversas caracteristicas fisicas. Asi mismo, se solucionan ejercicios ilustrativos a lo largo del articulo.

Control structures design for fossil power station units; Diseno de estructuras de control para unidades termoelectricas

Energy Technology Data Exchange (ETDEWEB)

Palomares Gonzalez, Daniel; Ricano Castillo, Juan Manuel [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1990-12-31

In this paper are designed an analyzed the control ties of a 300MW fossil unit at different operation levels. The design and analysis were developed on an operation range of 50% to 100% load. The main control ties were studied among which are the feedwater flow, the superheating zone temperatures and the generated output. The technique used for the design is the decomposition in singular values (DSV) of the plant transference matrix. This is a recent technique that permits finding the most important interactions among output-input variables, over a significative frequencies range for the transference matrix of the plant. For this purpose utilizes the control properties of the transference matrix, such as: control effort, output-input interaction and ruggedness with respect to uncertainties. The control structures obtained were tested in a digital model of a fossil plant unit. A comparison study of both structures (original and modified) showed enhancement in the performance of the power plant with modified structure. [Espanol] En este trabajo se disenan y analizan los lazos de control de una unidad termoelectrica de 300 MW, a diferentes niveles de operacion. El diseno y analisis se desarrollo sobre un rango de operacion de 50 a 100 por ciento de carga. Se estudiaron los principales lazos de control entre los que se encuentran el flujo de agua de alimentacion, las temperaturas de la zona de sobrecalentamiento y la potencia generada. La tecnica utilizada para el diseno es la descomposicion en valores singulares (DVS) de la matriz de transferencia de la planta. Esta es una tecnica reciente que permite encontrar las interacciones mas importantes entre variables entrada-salida, sobre un rango de frecuencias significativas para la matriz de transferencia de la planta. Utiliza para ello las propiedades de control de la matriz de transferencia, como son: esfuerzo de control, interaccion entrada-salida y robustez con respecto a incertidumbres. Las estructuras de control

Control structures design for fossil power station units; Diseno de estructuras de control para unidades termoelectricas

Energy Technology Data Exchange (ETDEWEB)

Palomares Gonzalez, Daniel; Ricano Castillo, Juan Manuel [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1989-12-31

In this paper are designed an analyzed the control ties of a 300MW fossil unit at different operation levels. The design and analysis were developed on an operation range of 50% to 100% load. The main control ties were studied among which are the feedwater flow, the superheating zone temperatures and the generated output. The technique used for the design is the decomposition in singular values (DSV) of the plant transference matrix. This is a recent technique that permits finding the most important interactions among output-input variables, over a significative frequencies range for the transference matrix of the plant. For this purpose utilizes the control properties of the transference matrix, such as: control effort, output-input interaction and ruggedness with respect to uncertainties. The control structures obtained were tested in a digital model of a fossil plant unit. A comparison study of both structures (original and modified) showed enhancement in the performance of the power plant with modified structure. [Espanol] En este trabajo se disenan y analizan los lazos de control de una unidad termoelectrica de 300 MW, a diferentes niveles de operacion. El diseno y analisis se desarrollo sobre un rango de operacion de 50 a 100 por ciento de carga. Se estudiaron los principales lazos de control entre los que se encuentran el flujo de agua de alimentacion, las temperaturas de la zona de sobrecalentamiento y la potencia generada. La tecnica utilizada para el diseno es la descomposicion en valores singulares (DVS) de la matriz de transferencia de la planta. Esta es una tecnica reciente que permite encontrar las interacciones mas importantes entre variables entrada-salida, sobre un rango de frecuencias significativas para la matriz de transferencia de la planta. Utiliza para ello las propiedades de control de la matriz de transferencia, como son: esfuerzo de control, interaccion entrada-salida y robustez con respecto a incertidumbres. Las estructuras de control

Conceptual design study for a laser fusion hybrid

International Nuclear Information System (INIS)

Maniscalco, J.A.

1976-01-01

Lawrence Livermore Laboratory and Bechtel Corporation have been involved in a joint effort to conceptually design a laser fusion hybrid reactor. The design which has evolved is a depleted-uranium fueled fast-fission blanket which produces fissile plutonium and electricity. A major objective of the design study was to evaluate the feasibility of producing fissile fuel with laser fusion. This feasibility evaluation was carried out by analyzing the integrated engineering performance of the complete conceptual design and by identifying the required laser/pellet performance. The performance of the laser fusion hybrid has also been compared to a typical fast breeder reactor. The results show that the laser fusion hybrid produces enough fissile material to fuel more than six light water reactors (LWRs) of equivalent thermal power while operating in a regime which requires an order of magnitude less laser and pellet performance than pure laser fusion. In comparison to a fast breeder reactor the hybrid produces 10 times more fissile fuel. An economic analysis of the design shows that the cost of electricity in a combined hybrid-LWR scenario increases by only 20 to 40 percent when the capital cost of the hybrid ranges from 2 to 3 times more than an LWR

Conceptual design study for a laser fusion hybrid

International Nuclear Information System (INIS)

Maniscalco, J.A.

1976-09-01

Lawrence Livermore Laboratory and Bechtel Corporation have been involved in a joint effort to conceptually design a laser fusion hybrid reactor. The design which has evolved is a depleted-uranium fueled fast-fission blanket which produces fissile plutonium and electricity. A major objective of the design study was to evaluate the feasibility of producing fissile fuel with laser fusion. This feasibility evaluation was carried out by analyzing the integrated engineering performance of the complete conceptual design and by identifying the required laser/pellet performance. The performance of the laser fusion hybrid has also been compared to a typical fast breeder reactor. The results show that the laser fusion hybrid produces enough fissile material to fuel more than six light water reactors (LWR's) of equivalent thermal power while operating in a regime which requires an order of magnitude less laser and pellet performance than pure laser fusion. In comparison to a fast breeder reactor the hybrid produces 10 times more fissile fuel. An economic analysis of the design shows that the cost of electricity in a combined hybrid-LWR scenario is insensitive to the capital cost of the hybrid, increasing by only 20 to 40 percent when the capital cost of the hybrid ranges from 2 to 3 times more than an LWR

Conceptual Design Tool for Fuel-Cell Powered Micro Air Vehicles

Science.gov (United States)

2010-03-01

Electrolyte Membrane PEMFC PEM Fuel Cell RAM Rapid Aircraft Modeler R/C Radio Controlled RMFC Reformed Methanol Fuel Cell SBIR Small Business...of rechargeable batteries, the Proton Exchange Membrane Fuel Cell ( PEMFC ) is only limited by the amount of hydrogen it can store, and can be...of fuel cells within MAVs through the creation of the Hornet. This slightly heavier, 380 g MAV integrated a 10 W PEMFC into the wing surface for a

Software for the design of acoustical steam silencers of the reaction-absorption type; Software para el diseno de silenciadores acusticos de vapor del tipo reaccion-absorcion

Energy Technology Data Exchange (ETDEWEB)

Buendia Dominguez, Eduardo H; Alvarez Chavez, Jose M [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1990-12-31

This paper describes the computer program named SILRA, that determines the outline dimensions of an acoustic steam silencer of the reaction-absorption type. These silencers are employed in the geothermoelectric power plants to lower the high levels of pressure and sound caused by the steam discharge to the surrounding atmosphere. The program has the capacity of predicting the noise level generated by the discharge without silencer depending on the emitting source. On SIRLA was successfully coupled the theory described in specialized literature with optimization techniques and experiences acquired in former designs. SIRLA is a powerful tool that allows the designer to optimize the equipment as well as the design time. [Espanol] En este trabajo se describe el programa de computo SILRA, que determina las dimensiones generales de un silenciador acustico de vapor tipo reaccion-absorcion, estos silenciadores se emplean en las centrales geotermoelectricas para abatir los altos niveles de presion de sonido provocados por la descarga del vapor a la atmosfera. El programa tiene la capacidad de predecir el nivel de ruido que genera la descarga sin silenciador, dependiendo de la fuente emisora. En SILRA se acoplaron con exito la teoria descrita en la literatura especializada con tecnicas de optimacion y experiencias adquiridas en disenos anteriores. SILRA es una poderosa herramienta que permite al disenador optimar tanto el equipo como el tiempo de diseno.

Software for the design of acoustical steam silencers of the reaction-absorption type; Software para el diseno de silenciadores acusticos de vapor del tipo reaccion-absorcion

Energy Technology Data Exchange (ETDEWEB)

Buendia Dominguez, Eduardo H.; Alvarez Chavez, Jose M. [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1989-12-31

This paper describes the computer program named SILRA, that determines the outline dimensions of an acoustic steam silencer of the reaction-absorption type. These silencers are employed in the geothermoelectric power plants to lower the high levels of pressure and sound caused by the steam discharge to the surrounding atmosphere. The program has the capacity of predicting the noise level generated by the discharge without silencer depending on the emitting source. On SIRLA was successfully coupled the theory described in specialized literature with optimization techniques and experiences acquired in former designs. SIRLA is a powerful tool that allows the designer to optimize the equipment as well as the design time. [Espanol] En este trabajo se describe el programa de computo SILRA, que determina las dimensiones generales de un silenciador acustico de vapor tipo reaccion-absorcion, estos silenciadores se emplean en las centrales geotermoelectricas para abatir los altos niveles de presion de sonido provocados por la descarga del vapor a la atmosfera. El programa tiene la capacidad de predecir el nivel de ruido que genera la descarga sin silenciador, dependiendo de la fuente emisora. En SILRA se acoplaron con exito la teoria descrita en la literatura especializada con tecnicas de optimacion y experiencias adquiridas en disenos anteriores. SILRA es una poderosa herramienta que permite al disenador optimar tanto el equipo como el tiempo de diseno.

Conceptual cask design with burnup credit

International Nuclear Information System (INIS)

Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong

2003-01-01

Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)

Overview of the CANDU fuel handling system for advanced fuel cycles

International Nuclear Information System (INIS)

Koivisto, D.J.; Brown, D.R.

1997-01-01

Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

Mirror hybrid reactor blanket and power conversion system conceptual design

International Nuclear Information System (INIS)

Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

1976-01-01

The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

Technical study report on fuel fabrication system

International Nuclear Information System (INIS)

Kono, Shusaku; Tanaka, Kenya; Ono, Kiyoshi; Iwasa, Katsuyoshi; Hoshino, Yasushi; Shinkai, Yasuo

2000-07-01

The feasibility study of FBR and related fuel cycle is performed for developing the FBR recycle system which ensures safety, economic competitiveness, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation under consistency of FBR reactor and fuel cycle systems. In this study, a conceptual design study and system characteristics evaluation are conducted for fuel fabrication systems of pellet process, vibropack process for oxide and nitride fuel and casting process for metal fuel. Technical issues in each process are also extracted. In 1999 fiscal year, a conceptual design study were conducted for the fuel fabrication plants adopting (1) the short pellet process which simplifies the conventional MOX pellet fabrication processes, (2) vibropack processes of aqueous gelation process, improved RIAR process, improved ANL process and fluoride volatility process, (3) casting processes of injection process, centrifuging process. As a result, attainable perspective was obtained for each fuel fabrication system through the evaluation of apparatuses, layout and facility volume, etc. In each fuel fabrication system, technical issues for practical use were made clear. Hereafter, more detailed study will be performed for each system, and research programs for phase II study will be planned. (author)

Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

Energy Technology Data Exchange (ETDEWEB)

IJ van Rooyen; SR Morrell; AE Wright; E. P Luther; K Jamison; AL Crawford; HT III Hartman

2014-10-01

Resumption of transient testing at the TREAT facility was approved in February 2014 to meet U.S. Department of Energy (DOE) objectives. The National Nuclear Security Administration’s Global Threat Reduction Initiative Convert Program is evaluating conversion of TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU). This paper describes briefly the initial pre-conceptual designs screening decisions with more detailed discussions on current feasibility, qualification and fabrication approaches. Feasible fabrication will be shown for a LEU fuel element assembly that can meet TREAT design, performance, and safety requirements. The statement of feasibility recognizes that further development, analysis, and testing must be completed to refine the conceptual design. Engineering challenges such as cladding oxidation, high temperature material properties, and fuel block fabrication along with neutronics performance, will be highlighted. Preliminary engineering and supply chain evaluation provided confidence that the conceptual designs can be achieved.

Progress on the conceptual design of a mirror hybrid fusion--fission reactor

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Burleigh, R.J.

1975-01-01

A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

International Nuclear Information System (INIS)

Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

2001-01-01

A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

Fuel Transfer Cask; Procedure Option and Radiation Protection during Transferring the Spent Fuel

International Nuclear Information System (INIS)

Muhammad Khairul Ariff Mustafa; Nurhayati Ramli; Ahmad Nabil Abdul Rahim; Mohd Fazli Zakaria

2011-01-01

Reactor TRIGA PUSPATI (RTP) has been operating almost 30 years. Many components are ageing. Nuclear Malaysia has taken an initiative to manage this ageing problem to prolong the life of the reactor. Hence, reactor upgrading project already commence started with the reactor console. To upgrade the core, all the fuel must be taken out from the core. A conceptual design of fuel transfer cask already done. This paper will discuss about the option of safe working procedure for transferring the fuel to the spent fuel pool for temporary. Hence, radiation protection for operator should be considered during the process. (author)

Metallic fuel design development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual design technology on metallic fuel

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

Energy Technology Data Exchange (ETDEWEB)

Diamond, David

2017-12-29

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allow for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.

General-purpose heat source development. Phase II: conceptual designs

International Nuclear Information System (INIS)

Snow, E.C.; Zocher, R.W.; Grinberg, I.M.; Hulbert, L.E.

1978-11-01

Basic geometric module shapes and fuel arrays were studied to determine how well they could be expected to meet the General Purpose Heat Source (GPHS) design requirements. Seven conceptual designs were selected, detailed drawings produced, and these seven concepts analyzed. Three of these design concepts were selected as GPHS Trial Designs to be reanalyzed in more detail and tested. The geometric studies leading to the selection of the seven conceptual designs, the analyses of these designs, and the selection of the three trial designs are discussed

Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

International Nuclear Information System (INIS)

1978-09-01

Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

Conceptual design of the special nuclear material nondestructive assay and accountability system for the HTGR fuel refabrication pilot plant

International Nuclear Information System (INIS)

Jenkins, J.D.; McNeany, S.R.; Rushton, J.E.

1975-07-01

The conceptual design of the fissile material assay and accountability system for the HTGR refabrication pilot plant has been established. The primary feature affecting the design is the high, time varying, gamma activity of the process material due to the unavoidable presence of uranium-232. This imposes stringent requirements for remote operation and remote maintainability of system components. At the same time, the remote operation lends itself to implementation of an automated data collection and processing system for real-time accountability. The high time-varying gamma activity of the material also precludes application of a number of techniques presently employed for light-water reactor fuel assay. The techniques selected for application in the refabrication facility are (1) active thermal neutron interrogation with fast-fission or delayed-neutron counting for fuel-rod and small-sample assay, (2) calorimetry for high-level waste assay, and (3) passive gamma scanning for low-level waste assay, and rapid on-line relative rod-loading measurements. The principal nondestructive assay subsystems are identified as (1) on-line devices for 100 percent product fuel rod assay and quality control, (2) a multipurpose device in the sample inspection laboratory for small- sample assay and secondary standards calibration, and (3) equipment for assay of high- and low-uranium content scrap and waste materials. A data processing system, which coordinates data from these subsystems with information from other process control sensors, is included to provide real-time material balance information. (U.S.)

Development of nuclear fuel. Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

1991-07-01

In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

Flow analysis of tubular fuel assembly using CFD code

International Nuclear Information System (INIS)

Park, J. H.; Park, C.; Chae, H. T.

2004-01-01

Based on the experiences of HANARO, a new research reactor is under conceptual design preparing for future needs of research reactor. Considering various aspects such as nuclear physics, thermal-hydraulics, mechanical structure and the applicability of HANARO technology, a tubular type fuel has been considered as that of a new research reactor. Tubular type fuel has several circular fuel layers, and each layer consists of 3 curved fuel plates arranged with constant small gap to build up cooling channels. In the thermal-hydraulic point, it is very important to maintain each channel flow velocity be equal as much as possible, because the small gaps between curved thin fuel plates independently forms separate coolant channels, which may cause a thermal-hydraulic problem in certain conditions. In this study, commercial CFD(Computational Fluid Dynamics) code, Fluent, has been used to investigate flow characteristics of tubular type fuel assembly. According to the computation results for the preliminary conceptual design, there is a serious lack of uniformity of average velocity on the each coolant channel. Some changes for initial conceptual design were done to improve the balance of velocity distribution, and analysis was done again, too. The results for the revised design showed that the uniformity of each channel velocity was improved significantly. The influence of outermost channel gap width on the velocity distribution was also examined

Conceptual design of next generation MTR

Energy Technology Data Exchange (ETDEWEB)

Nagata, Hiroshi; Yamaura, Takayuki; Naka, Michihiro; Kawamata, Kazuo; Izumo, Hironobu; Hori, Naohiko; Nagao, Yoshiharu; Kusunoki, Tsuyoshi; Kaminaga, Masanori; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Mine, M [Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki (Japan); Yamazaki, S [Kawasaki Heavy Industries, Ltd., Kobe, Hyogo (Japan); Ishikawa, S [NGK Insulators, Ltd., Nagoya, Aichi (Japan); Miura, K [Sukegawa Electric Co., Ltd., Takahagi, Ibaraki (Japan); Nakashima, S [Fuji Electric Co., Ltd., Tokyo (Japan); Yamaguchi, K [Chiyoda Technol Corp., Tokyo (Japan)

2012-03-15

Conceptual design of the high-performance and low-cost next generation materials testing reactor (MTR) which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

System to solve three designs of the fuel management; Sistema para resolver tres disenos de la administracion de combustible

Energy Technology Data Exchange (ETDEWEB)

Castillo M, J. A.; Ortiz S, J. J.; Montes T, J. L.; Perusquia del C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Marinez R, R., E-mail: alejandro.castillo@inin.gob.mx [Universidad Autonoma de Campeche, Av. Agustin Melgar s/n, Col. Buenavista, 24039 San Francisco de Campeche, Campeche (Mexico)

2015-09-15

In this paper preliminary results are presented, obtained with the development of a computer system that resolves three stages of the nuclear fuel management, which are: the axial and radial designs of fuel, as well as the design of nuclear fuel reloads. The novelty of the system is that the solution is obtained solving the 3 mentioned stages, in coupled form. For this, heuristic techniques are used for each stage, in each one of these has a function objective that is applied to particular problems, but in all cases the obtained partial results are used as input data for the next stage. The heuristic techniques that were used to solve the coupled problem are: tabu search, neural networks and a hybrid between the scatter search and path re linking. The system applies an iterative process from the design of a fuel cell to the reload design, since are preliminary results the reload is designed using the operation strategy Haling type. In each one of the stages nuclear parameters inherent to the design are monitored. The results so far show the advantage of solving the problem in a coupled manner, even when a large amount of computer resources is used. (Author)

Development of on-site spent fuel transfer system designs

International Nuclear Information System (INIS)

Lambert, R.W.; Pennington, C.W.; Guerra, G.V.

1993-01-01

The Electric Power Research Institute (EPRI) of the United States has sponsored development of conceptual designs for accomplishing spent fuel transfer from spent fuel pools to casks and from one cask to another. Under an EPRI research contract, transnuclear has developed several concepts for spent fuel transfer systems. (J.P.N.)

Pirotubular boilers design. Diseno de calderas pirotubulares

Energy Technology Data Exchange (ETDEWEB)

Latre Durso, F. (Geval S.A. (Spain))

1994-01-01

This article describes the conceptual design of Pirotubular boilers from the valuable thermal point of view, to the dimensional. This topic is a very tool valuable for professionals of design and for maintenance and operation equipment.

Development of a fresh MOX fuel transport package for disposition of weapons plutonium

International Nuclear Information System (INIS)

Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D.; Chae, S.M.

1998-01-01

The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements

Design of a new research reactor : 1st year conceptual design

International Nuclear Information System (INIS)

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

2004-01-01

A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

Conceptual study

Energy Technology Data Exchange (ETDEWEB)

Harty, H.

1978-09-01

This appendix is a compendium of topical reports prepared for the Hanford Nuclear Energy Center: Status Report: Conceptual Fuel Cycle Studies for the Hanford Nuclear Energy Center; Selection of Heat Disposal Methods for a Hanford Nuclear Energy Center; Station Service Power Supply for a Hanford Nuclear Energy Center (HNEC); Impact of a Hanford Nuclear Energy Center on Ground Level Fog and Humidity; A Review of Potential Technology for the Seismic Characterization of Nuclear Energy Centers; Reliability of Generation at a Hanford Nuclear Energy Center (HNEC); Meteorological Evaluation of Multiple Reactor Contamination Probabilities for a Hanford Nuclear Energy Center; Electric Power Transmission for a Hanford Nuclear Energy Center (HNEC); The Impact of a Hanford Nuclear Energy Center on Cloudiness and Insolation; and A Licensing Review for an HNEC.

Conceptual study

International Nuclear Information System (INIS)

Harty, H.

1978-09-01

This appendix is a compendium of topical reports prepared for the Hanford Nuclear Energy Center: Status Report: Conceptual Fuel Cycle Studies for the Hanford Nuclear Energy Center; Selection of Heat Disposal Methods for a Hanford Nuclear Energy Center; Station Service Power Supply for a Hanford Nuclear Energy Center (HNEC); Impact of a Hanford Nuclear Energy Center on Ground Level Fog and Humidity; A Review of Potential Technology for the Seismic Characterization of Nuclear Energy Centers; Reliability of Generation at a Hanford Nuclear Energy Center (HNEC); Meteorological Evaluation of Multiple Reactor Contamination Probabilities for a Hanford Nuclear Energy Center; Electric Power Transmission for a Hanford Nuclear Energy Center (HNEC); The Impact of a Hanford Nuclear Energy Center on Cloudiness and Insolation; and A Licensing Review for an HNEC

A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

International Nuclear Information System (INIS)

Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

2009-01-01

The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

Metallic fuel design development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

Conceptual design of a spent LWR fuel recycle complex

International Nuclear Information System (INIS)

Kirk, B.H.

1980-01-01

Purpose was to design a licensable facility, to make cost-benefit analyses of alternatives, and to aid in developing licensing criteria. The Savannah River Plant was taken to be the site for the recycle complex. The spent LWR fuel will be processed through the plant at the rate of 3000 metric tons of heavy metal per year. The following aspects of the complex are discussed: operation, maintenance, co-conversion (Coprecal), waste disposal, off-gas treatment, ventilation, safeguards, accounting, equipment and fuel fabrication. Differences between the co-processing case and the separated streams case are discussed. 44 figures

Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

Energy Technology Data Exchange (ETDEWEB)

Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2011-11-15

At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

International Nuclear Information System (INIS)

Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

2012-06-01

The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

Conceptual design of multipurpose compact research reactor

International Nuclear Information System (INIS)

Nagata, Hiroshi; Kusunoki, Tsuyoshi; Hori, Naohiko; Kaminaga, Masanori

2012-01-01

Conceptual design of the high-performance and low-cost multipurpose compact research reactor which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

Radionuclides and isotopes release of spent fuel matrix. Conceptual and mathematical models of wastes behaviour

International Nuclear Information System (INIS)

Cera, E.; Merino, J.; Bruno, J.

2000-01-01

We have developed a conceptual and numerical model to calculate release of selected radionuclides from spent fuel under repository condition. This has been done in the framework of the Enresa 2000 performance assessment exercise. The model has been developed based on kinetic mass balance equations in order to study the evolution of the spent fuel water interface as a function of time. Several processes have been kinetically modelled: congruent dissolution, radioactive decay, ingrowth and water turnover in the gap. The precipitation/redissolution of secondary solid phases has been taken into account from a thermodynamic point of view. Both approaches have been coupled and the resulting equations solved for a number of radionuclides in both, a conservative and realistic approach. The results show three distinct groups of radionuclides based on their release behaviour: a first group is composed of radioisotopes of highly insoluble elements (e. g., Pu, Am, Pd) whose concentration in the gap is mainly controlled by their solubility and therefore their evolution is identical in both cases. Secondly, a set of radionuclides from soluble elements under these conditions (e. g., I, Cs, Ra) show concentrations kinetically controlled, decreasing with time following the congruent dissolution trend. Their release concentrations are one order of magnitude larger in the conservative case than in the realistic case. Finally, a third group has been identified (e. g., Se, Th, Cm) where a mixed behaviour takes place: initially their solubility limiting phases control their concentration in the gap but the situation reverts to a kinetic control as the chemical conditions change and the secondary precipitates become totally dissolved. The fluxes of the different radionuclides are also given as an assessment of the source term in the performance assessment. (Author)

Conceptual designing of reduced-moderation water reactor with heavy water coolant

Energy Technology Data Exchange (ETDEWEB)

Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

2001-12-01

The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

Issues for Conceptual Design of AFCF and CFTC LWR Spent Fuel Separations Influencing Next-Generation Aqueous Fuel Reprocessing

International Nuclear Information System (INIS)

D. Hebditch; R. Henry; M. Goff; K. Pasamehmetoglu; D. Ostby

2007-01-01

In 2007, the U.S. Department of Energy (DOE) published the Global Nuclear Energy Partnership (GNEP) strategic plan, which aims to meet US and international energy, safeguards, fuel supply and environmental needs by harnessing national laboratory R and D, deployment by industry and use of international partnerships. Initially, two industry-led commercial scale facilities, an advanced burner reactor (ABR) and a consolidated fuel treatment center (CFTC), and one developmental facility, an advanced fuel cycle facility (AFCF) are proposed. The national laboratories will lead the AFCF to provide an internationally recognized R and D center of excellence for developing transmutation fuels and targets and advancing fuel cycle reprocessing technology using aqueous and pyrochemical methods. The design drivers for AFCF and the CFTC LWR spent fuel separations are expected to impact on and partly reflect those for industry, which is engaging with DOE in studies for CFTC and ABR through the recent GNEP funding opportunity announcement (FOA). The paper summarizes the state-of-the-art of aqueous reprocessing, gives an assessment of engineering drivers for U.S. aqueous processing facilities, examines historic plant capital costs and provides conclusions with a view to influencing design of next-generation fuel reprocessing plants

Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

2000-08-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect.

Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

International Nuclear Information System (INIS)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

2000-08-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect

Conceptual designs of radioactive canister transporters

International Nuclear Information System (INIS)

1978-02-01

This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts

Nanofluidic fuel cell

Science.gov (United States)

Lee, Jin Wook; Kjeang, Erik

2013-11-01

Fuel cells are gaining momentum as a critical component in the renewable energy mix for stationary, transportation, and portable power applications. State-of-the-art fuel cell technology benefits greatly from nanotechnology applied to nanostructured membranes, catalysts, and electrodes. However, the potential of utilizing nanofluidics for fuel cells has not yet been explored, despite the significant opportunity of harnessing rapid nanoscale reactant transport in close proximity to the reactive sites. In the present article, a nanofluidic fuel cell that utilizes fluid flow through nanoporous media is conceptualized and demonstrated for the first time. This transformative concept captures the advantages of recently developed membraneless and catalyst-free fuel cell architectures paired with the enhanced interfacial contact area enabled by nanofluidics. When compared to previously reported microfluidic fuel cells, the prototype nanofluidic fuel cell demonstrates increased surface area, reduced activation overpotential, superior kinetic characteristics, and moderately enhanced fuel cell performance in the high cell voltage regime with up to 14% higher power density. However, the expected mass transport benefits in the high current density regime were constrained by high ohmic cell resistance, which could likely be resolved through future optimization studies.

Conceptual design of interim storage facility for CNAI

International Nuclear Information System (INIS)

Fuenzalida Troyano, Carlos S.; Bergallo, Juan E.; Nassini, Horacio E.P.; Blanco, Anibal; Delmastro, Dario F.

2007-01-01

The reduced storage capacity available in the two spent fuel pools of argentine PHWR Atucha-1 power plant, the current plans for extending the reactor operation beyond its design lifetime, and the government decision on Atucha-2 NPP construction ending, have motivated the evaluation of a dry storage option for the interim management of spent fuel assemblies. Two different designs are presently being analyzed by an expert working group, from both technical and economical points of views. Authors are proposing a modular system consisting of an arrangement of reinforced concrete structures into which welded metallic canisters loaded with 37 spent fuel assemblies each stored in horizontal position. The reinforced concrete module is designed to provide the necessary physical protection and biological shielding to the loaded canisters during long-term storage, as well as passive means to remove the spent fuel decay heat by a combination of radiation, conduction and natural air convection. In this works are presented advances in the conceptual designs for a spent nuclear fuel system to Atucha I nuclear power plant. (author) [es

Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

International Nuclear Information System (INIS)

Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

1990-02-01

A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

300 kWt core conceptual model thermal/hydraulic characteristics

International Nuclear Information System (INIS)

Moody, E.

1971-01-01

The 300 kW(t), 199 element NASA-Lewis/AEC core conceptual model, has been analyzed to determine it's thermal-hydraulic characteristics using the GEOM-3 code. Stack-ups of tolerances and fuel rod asymmetry patterns were used to ascertain cross element Δ T's. Both zoned and uniform spacing were analyzed with a somewhat lower fuel temperature and cross element ΔT found for zoned spacing. With the models considered, the core design appears adequate to limit thermal gradients to approximately 32 0 F. Bypass flow should be controlled to prevent excessive edge element ΔT's. 11 references. (U.S.)

Design and construction of an electrolyte PEM test; Diseno y construccion de un electrolizador PEM de prueba

Energy Technology Data Exchange (ETDEWEB)

Gonzalez-Huerta, R. G.; Santillan-Aragon, G. [Instituto Politecnico Nacional, Mexico, D.F. (Mexico)]. E-mail: rosgonzalez_h@yahoo.com.mx; Solorza-Feria, O. [CINVESTAV-IPN, Mexico D.F. (Mexico)

2009-09-15

The use of hydrogen as a fuel is directly linked to its efficient and clean production. One of the most promising methods is water electrolysis, which coupled with a renewable energy source prevents the emission of pollutants into the atmosphere. If a proton exchange membrane (PEM) electrolysis is used, a highly pure hydrogen is produced, ready to be used in a fuel battery. Many studies and investigations in this area concentrate on finding different stable and selective electrocatalysts for the cathode reaction (production of hydrogen) and anode reaction (production of oxygen). To conduct these studies, equipment is needed to perform electrochemical studies and determine the stability and performance of different electrocatalysts. This work presents the design and construction of an a PEM electrolysis test to determine the performance of different anode electrocatalysts. Its active area is 4 cm{sup 2}, its structure is graphite and the current distribution mesh is made of stainless steel. Its performance was determined using as electrocatalysts 10% Pt/C E-tek® anodes and a 50%-50%, 25%-75% and 75%-25% combination of RuO{sub 2}-IrO{sub 2}. The authors wish to thank the ICYTDF (PICS08-37) for financial support and IPN (SIP-20090433) and architect Nestor Romero for the electrolysis machining. [Spanish] La utilizacion del hidrogeno como combustible esta ligado directamente a su produccion eficiente y limpia, uno de los metodos mas prometedores es la electrolisis del agua, ya que acoplado con una fuente de energia renovable se evita la emision de contaminantes a la atmosfera. Si se utiliza un electrolizador de membrana de intercambio protonico (Tipo PEM), el hidrogeno que se produce es de alta pureza, listo para ser utilizado en una pila de combustible. Muchos estudios e investigaciones en esta area se concentran en encontrar distintos electrocatalizadores estables y selectivos para la reaccion catodica (produccion de hidrogeno) y anodica (produccion de oxigeno). Para

Engineered safeguards system activities at Sandia Laboratories for back-end fuel cycle facilities

International Nuclear Information System (INIS)

Sellers, T.A.; Fienning, W.C.; Winblad, A.E.

1978-01-01

Sandia Laboratories have been developing concepts for safeguards systems to protect facilities in the back-end of the nuclear fuel cycle against potential threats of sabotage and theft of special nuclear material (SNM). Conceptual designs for Engineered Safeguards Systems (ESSs) have been developed for a Fuel Reprocessing Facility (including chemical separations, plutonium conversion, and waste solidification), a Mixed-Oxide Fuel Fabrication Facility, and a Plutonium Transport Vehicle. Performance criteria for the various elements of these systems and a candidate systematic design approach have been defined. In addition, a conceptual layout for a large-scale Fuel-Cycle Plutonium Storage Facility has been completed. Work is continuing to develop safeguards systems for spent fuel facilities, light-water reactors, alternative fuel cycles, and improved transportation systems. Additional emphasis will be placed on the problems associated with national diversion of special nuclear material. The impact on safeguards element performance criteria for surveillance and containment to protect against national diversion in various alternative fuel cycle complexes is also being investigated

Conceptual design study of high temperature gas-cooled reactor for plutonium incineration

International Nuclear Information System (INIS)

Goto, Minoru

2013-01-01

JAEA has started a conceptual design study of a Pu burner HTGR, which is called CBHTR (Clean Burn High Temperature gas-cooled Reactor). CBHTR’s fuel is TRISO-coated fuel particle with PuO 2 -YSZ (Yttria- Stabilized Zirconia) kernel, which increase proliferation resistance, safety of geological disposal, and Pu incineration. CBHTR can decrease Puf ratio from 60% to 20% with 520 GWd/t. In the future, 15% of electricity capacity is employed by 7 of CBHTRs and 59 of U-HTRs. JAEA has a R and D plan of manufacturing technology of TRISO-coated fuel with PuO 2 -YSZ kernel

Actual Status of CAREM-25 Fuel Element Development

International Nuclear Information System (INIS)

Perez, Edmundo

2000-01-01

In the frame of the CAREM Project, under Cnea s Reactor and Nuclear Plants Program, the Nuclear Fuel Thematic Area is one among others on which the project is organized. In this area, the primary objective to reach is to actualize the mechanical fuel element and reactivity control designs, taking in account the recents conceptual and engineering modifications introduced in the reactor, and ending with a consolidated conceptual and basic development.In order to reach these objectives, it is presented the way on which the area was organized, the participating working groups, the task required, the personnel involucrated, the grade of global development reached in the areas of engineering, developments, fabrication and essays of design verification, and the found difficulties, the tasks under ejecution, just finished and necessaries to fulfill completely the objectives. Finally, it is possible to say that due to the work realized, the conceptual design of both components is finished and the basic design is under development

Conceptos de diseño para manufactura (DFM de piezas microfundidas

Directory of Open Access Journals (Sweden)

Carlos Julio Cortés Rodríguez

2005-09-01

Full Text Available Una de las metodologIas utilizadas por la ingenierIa concurrente o diseno integrado de productos es el diseno para manufactura, DEM. Esta técnica, que está orientada a melorar la fabricación de piezas analizando geometrIas, valores y tolerancias, es utilizada en las primeras etapas de diseno para disminuir los costos de los elementos a fabricar. Se estudió el diseno para manufactura estableciendo reglas de diseno a fin de elaborar una metodologIa para diseno en microfundición.

Monitoring instrumentation spent fuel management program. Final report

International Nuclear Information System (INIS)

1979-01-01

Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated

Fuel Flexible Combustion Systems for High-Efficiency Utilization of Opportunity Fuels in Gas Turbines

Energy Technology Data Exchange (ETDEWEB)

Venkatesan, Krishna

2011-11-30

The purpose of this program was to develop low-emissions, efficient fuel-flexible combustion technology which enables operation of a given gas turbine on a wider range of opportunity fuels that lie outside of current natural gas-centered fuel specifications. The program encompasses a selection of important, representative fuels of opportunity for gas turbines with widely varying fundamental properties of combustion. The research program covers conceptual and detailed combustor design, fabrication, and testing of retrofitable and/or novel fuel-flexible gas turbine combustor hardware, specifically advanced fuel nozzle technology, at full-scale gas turbine combustor conditions. This project was performed over the period of October 2008 through September 2011 under Cooperative Agreement DE-FC26-08NT05868 for the U.S. Department of Energy/National Energy Technology Laboratory (USDOE/NETL) entitled "Fuel Flexible Combustion Systems for High-Efficiency Utilization of Opportunity Fuels in Gas Turbines". The overall objective of this program was met with great success. GE was able to successfully demonstrate the operability of two fuel-flexible combustion nozzles over a wide range of opportunity fuels at heavy-duty gas turbine conditions while meeting emissions goals. The GE MS6000B ("6B") gas turbine engine was chosen as the target platform for new fuel-flexible premixer development. Comprehensive conceptual design and analysis of new fuel-flexible premixing nozzles were undertaken. Gas turbine cycle models and detailed flow network models of the combustor provide the premixer conditions (temperature, pressure, pressure drops, velocities, and air flow splits) and illustrate the impact of widely varying fuel flow rates on the combustor. Detailed chemical kinetic mechanisms were employed to compare some fundamental combustion characteristics of the target fuels, including flame speeds and lean blow-out behavior. Perfectly premixed combustion experiments were conducted to

Conversion of fuel hulls to zirconate ion exchangers for stabilization of wastes from the thorium fuel cycle

International Nuclear Information System (INIS)

Levine, H.S.

1978-01-01

A conceptual reprocessing and waste management scheme for Zircaloy clad ThO 2 fuel was formulated to eliminate problems associated with concurrent dissolution of fuel and cladding in the conventional chop-leach headend step. These problems are avoided by use of a modified headend step to form oxide fuel and cladding process streams. A chlorinating agent then converts the cladding hulls and adhering fuel into volatile and nonvolatile chloride fractions. The former product is processed, by use of the Zircaloy conversion process, to form an inorganic ion exchange material and combined with HLLW from subsequent fuel reprocessing to form a stable and refractory waste form. The nonvolatile chloride fraction would be recovered, processed to remove chloride ions, and recombined with the main oxide fuel process stream for further treatment by use of the Thorex process

Irradiation experiment conceptual design parameters for MURR LEU U-Mo fuel conversion

International Nuclear Information System (INIS)

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

2013-03-01

This report contains the results of reactor design and performance calculations for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the nominal steady-state irradiation conditions of a key set of plates containing peak irradiation parameters found in MURR cores fueled with the LEU monolithic U-Mo alloy fuel with 10 wt% Mo.

The System Dynamics of U.S. Automobile Fuel Economy

Directory of Open Access Journals (Sweden)

Todd K. BenDor

2012-05-01

Full Text Available This paper analyzes the dynamics of U.S. automobile gasoline consumption since 1975. Using background literature on the history of domestic fuel economy and energy policy, I establish a conceptual model that explains historical trends in adoption of increased fuel economy. I then create a system dynamics simulation model to understand the relationship between increased fuel economy standards and potential changes to gas tax policies. The model suggests that when increases in mandated fuel economy are not conducted in an environment with rising fuel costs, fuel economy improvements may be directly counteracted by shifting tastes of consumers towards larger automobiles with lower fuel economy.

Waste Handling Building Conceptual Study

International Nuclear Information System (INIS)

G.W. Rowe

2000-01-01

The objective of the ''Waste Handling Building Conceptual Study'' is to develop proposed design requirements for the repository Waste Handling System in sufficient detail to allow the surface facility design to proceed to the License Application effort if the proposed requirements are approved by DOE. Proposed requirements were developed to further refine waste handling facility performance characteristics and design constraints with an emphasis on supporting modular construction, minimizing fuel inventory, and optimizing facility maintainability and dry handling operations. To meet this objective, this study attempts to provide an alternative design to the Site Recommendation design that is flexible, simple, reliable, and can be constructed in phases. The design concept will be input to the ''Modular Design/Construction and Operation Options Report'', which will address the overall program objectives and direction, including options and issues associated with transportation, the subsurface facility, and Total System Life Cycle Cost. This study (herein) is limited to the Waste Handling System and associated fuel staging system

Conceptual Design for the In-Pile Test Section(IPS) Improvement

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Min; Shim, Bong Sik; Lee, Chung Young; Chi, Dae Young; Park, Su Ki; Ahn, Sung Ho; Kim, Young Ki; Lee, Kye Hong; Kim, Kwan Hyun

2009-01-15

Conceptual design on the IPS, instrumentation and 1/4' Tubing, test fuel supports, test fuel inspection station and inner assembly O-ring replacement was suggested for their improvement. There is a need Jacking bolt on the Top flange for the inner assembly disassemble from the outer assembly and the replacement about the top flange O-ring to metal gasket to secure pressure boundary. Mechanical sealing was suggested instead of brazing. Instrumentation and tubing route should be modified for the reduction of bending and protection from unexpected occasion. Concept on the test fuel inspection station under the consideration of canal dimension. Top flange Bolt handling tool, O-ring replacement platform and O-ring replacement tool was designed for the O-ring replacement at the inner assembly.

Conceptual design of a nucleo electric simulator with PBMR reactor based in Reduced order models; Diseno conceptual de un simulador de nucleo electrica con reactor PBMR basado en modelos de orden reducido

Energy Technology Data Exchange (ETDEWEB)

Valle H, J.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

2005-07-01

This project has as purpose to know to depth the operation of a PBMR nucleo electric type (Pebble Bed Modular Reactor), which has a reactor of moderate graphite spheres and fuel of uranium dioxide cooled with Helium and Brayton thermodynamic cycle. The simulator seeks to describe the dynamics of the one process of energy generation in the nuclear fuel, the process of transport toward the coolant one and the conversion to mechanical energy in the turbo-generators as well as in the heat exchangers indispensable for the process. The dynamics of reload of the fuel elements it is not modeled in detail but their effects are represented in the parameters of the pattern. They are modeled also the turbo-compressors of the primary circuit of the work fluid. The control of the power of the nuclear reactor is modeled by means of reactivity functions specified in the simulation platform. The proposed mathematical models will be settled in the platform of simulation of Simulink-Mat Lab. The proposed control panels for this simulator can be designed and to implement using the box of tools of Simulink that facilitates this process. The work presents the mathematical models more important used for their future implementation in Simulink. (Author)

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

International Nuclear Information System (INIS)

1978-06-01

This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions

Plutonium fuel program

International Nuclear Information System (INIS)

1979-01-01

The work of the Project-Fuel Development reached the apex of its current programme during the course of the year. Notable success was recorded in the area of irradiation testing with the completion of the examination of the MFBS-7 irradiation. The irradiation group also prepared the seventh Filos experiment and this, as well as the DIDO-III test, began irradiation at the end of the year. Consideration was given and plans prepared for a revised pin filling line for bundle tests. Work also began on the conceptual design study for a pilot production line having a nominal capacity of 500 kg fuel per year. (Auth.)

Present status of fuel motion detection by radiation

International Nuclear Information System (INIS)

Sumita, Kenji; Mizuta, Hiroshi; Ishizuka, Makoto; Ara, Katsuyuki; Nakata, Hirokatsu.

1978-05-01

In reactor safety research, it is important to know transient fuel behavior under accidental conditions. Transient histories such as temperature and axial expansion of fuel and cladding and internal pressure of fuel rod are thus measured in experiments simulating accidents. If fuel motion could then be observed during and after fuel failure, this would greatly make for fuel behavior research. The present status is reviewed of fuel motion detections by radiations such as neutron, γ-ray and X-ray, including the principle and system. A neutron hodoscope among them is used already with practical results in in-reactor safety experiments of sodium-cooled fast breeder reactor. So, this is described in detail and its conceptual design as applied to the NSRR is presented. (auth.)

Conceptual design of Indian molten salt breeder reactor

International Nuclear Information System (INIS)

Vijayan, P.K.; Basak, A.; Dulera, I.V.; Vaze, K.K.; Basu, S.; Sinha, R.K.

2014-01-01

The fuel in a molten salt breeder reactor is in the form of a continuously circulating molten salt. Fluoride based salts have been almost universally proposed. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. This constitutes a major technological challenge for this type of reactors. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). Presently various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel fundamental studies as regards various molten salts have also been initiated. This paper would discuss conceptual design of these reactors, as well as associated issues and technologies

Studies and research concerning BNFP: spent fuel dry storage studies at the Barnwell Nuclear Fuel Plant

International Nuclear Information System (INIS)

Anderson, K.J.

1980-09-01

Conceptual designs are presented utilizing the Barnwell Nuclear Fuel Plant for the dry interim storage of spent light water reactor fuel. Studies were conducted to determine feasible approaches to storing spent fuel by methods other than wet pool storage. Fuel that has had an opportunity to cool for several years, or more, after discharge from a reactor is especially adaptable to dry storage since its thermal load is greatly reduced compared to the thermal load immediately following discharge. A thermal analysis was performed to help in determining the feasibility of various spent fuel dry storage concepts. Methods to reject the heat from dry storage are briefly discussed, which include both active and passive cooling systems. The storage modes reviewed include above and below ground caisson-type storage facilities and numerous variations of vault, or hot cell-type, storage facilities

Feasibility study for a DOE research and production fuel multipurpose canister

International Nuclear Information System (INIS)

Lopez, D.A.; Abbott, D.G.

1994-02-01

This is a report of the feasibility of multipurpose canisters for transporting, storing, and sing of Department of Energy research and production spent nuclear fuel. Six representative Department of Energy fuel assemblies were selected, and preconceptual canister designs were developed to accommodate these assemblies. The study considered physical interface, structural adequacy, criticality safety, shielding capability, thermal performance of the canisters, and fuel storage site infrastructure. The external envelope of the canisters was designed to fit within the overpack casks for commercial canisters being developed for the Department of Energy Office of Civilian Radioactive Waste Management. The budgetary cost of canisters to handle all fuel considered is estimated at $170.8M. One large conceptual boiling water reactor canister design, developed for the Office of Civilian Radioactive Waste Management, and two new canister designs can accommodate at least 85% of the volume of the Department of Energy fuel considered. Canister use minimizes public radiation exposure and is cost effective compared with bare fuel handling. Results suggest the need for additional study of issues affecting canister use and for conceptual design development of the three canisters

Solar system design to heating water for a biodigester; Diseno de un sistema solar de calentamiento de agua para un biodigestor

Energy Technology Data Exchange (ETDEWEB)

Gutierrez Castro, Lucia Monica; Quinto Diez, Pedro [Escuela Superior de Ingenieria Mecanica y Electrica, Instituto Politecnico Nacional, Mexico, D.F. (Mexico)]. E-mail: moni-80-lgc@hotmail.com; pqd510@hotmail.com; Tovar Galvez, Luis Raul [Centro Interdisciplinario de Investigaciones de Estudios sobre Medio Ambiente y Desarrollo, Instituto Politecnico Nacional, Mexico, D.F. (Mexico)]. E-mail: ltovarg@ipn.mx

2010-11-15

In this paper a Solar System of Heating Water (SSHW) to provide energy to a biodigester which produces biogas located in the Interdisciplinary Center of Investigation of Studies About Environment and Development (CIIEMAD) is assessed. This evaluation is made by means of the f Method considerating the extraterrestrial radiation, the earth radiation in the zone and the necessary energy to maintain at 55 degrees Celsius the mixture inside the biodigester. The data were introduced in the SOLAR program to calculate both the harnessing area m{sup 2} and the storage capacity in L/m{sup 2}. To guarantee the good performance of the SSHW, a control and monitoring system for the temperature of the mix inside the biodigester is proposed. It also controls the input and output temperature to the solar collectors, as well as the cost analysis which shows the savings of fuel of the SSHW with a conventional system to heating water. [Spanish] En este articulo se evalua un sistema solar de calentamiento de agua (SSCA) para proveer energia a un biodigestor que forma parte del sistema de generacion de biogas (SGB) ubicado en el Centro Interdisciplinario de Investigaciones de Estudios sobre Medio Ambiente y Desarrollo (CIIEMAD). Dicha evaluacion se realiza por medio del Metodo f de diseno que considera datos de radiacion terrestre, radiacion extraterrestre sobre la zona y la energia necesaria para mantener la temperatura de la mezcla dentro del biodigestor a 55 grados centigrados. Los datos fueron introducidos en el programa SOLAR para el calculo del area de captacion en m{sup 2} y la capacidad de almacenamiento en L/m{sup 2}. Para garantizar el buen funcionamiento del SSCA se propone un sistema de control y monitoreo de la temperatura de la mezcla dentro del biodigestor. Tambien se controla la temperatura de entrada y salida a los colectores solares, asi como el analisis de costos que muestra el ahorro de combustible del SSCA con un sistema convencional de calentamiento de agua.

Conceptual study of the future nuclear fuel cycle system for the extended LWR age

International Nuclear Information System (INIS)

Fujine, Sachio; Takano, Hideki; Sato, Osamu; Tone, Tatsuzo; Yamada, Takashi; Kurosawa, Katsutoshi.

1993-08-01

A large scale integrated fuel cycle facility (IFCF) is assumed for the future nuclear fuel cycle in the extended LWR age. Spent MOX fuels are reprocessed mixed with UOX in a centralized reprocessing plant. The reprocessing plant separates long-lived nuclides as well as Pu. Nitric acid solutions of those products are fed directly to MOX fabrication process which is incorporated with reprocessing. MOX pellets are made by sphere-cal process. Two process concepts are made as advanced reprocessing incorporated with partitioning (ARP) which has the function of long-lived nuclides recovery. One is a simplified Purex combined with partitioning. Extractable long-lived nuclides, 237 Np and 99 Tc, are assumed to be recovered in main flow stream of the improved Purex process. The other process concept is made aiming at recovering all TRU nuclides in reprocessing to meet with TRU recycle requirement in the long future. A concept of the future fuel cycle system is made by combining integrated fuel cycle facility and very high burnup LWRs (VHBR). The reactor concept of VHBRs has been proposed to improve Pu recycle economy in the future. Highly enriched MOX fuel are loaded in the full core of reactor in order to increase reactivity for the burnup. Fuel cycle indices such as Pu isotopic composition change, spent fuel integration, nuclide transmutation effect are estimated by simulating the Pu recycling in the system of VHBR and ARP. It is concluded that Pu enrichment of MOX fuel can be kept less than 20 % through multi-recycle. Reprocessing MOX fuels with UOX shows a favorable effect for keeping Pu reactivity high enough for VHBR. Integration of spent MOX fuel can be reduced by Pu recycle. Transmutation of Np is feasible by containing Np into MOX fuel. (author)

Fast reactor fuel design and development

International Nuclear Information System (INIS)

Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

1977-01-01

Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

Cost evaluation of a commercial-scale DUPIC fuel fabrication facility (Part I) -Summary

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-08-01

A conceptual design of a commercial scale DUPIC fuel fabrication facility was initiated to provide some insights into the costs associated with construction, operation, and decommissioning. The primary conclusion of this report is that it is feasible to design, license, construct, test, and operate a facility that will process 400 MTHE/yr of spent PWR fuel and reconfigure the fuel into CANDU fuel bundles at a reasonable unit cost of the fuel material. Although DUPIC fuel fabrication by vibropacking method is clearly cheaper than that of the pellet method, the feasibility of vibropac technology for DUPIC fuel fabrication and use of vibroac fuel in CANDU reactors may has to be studied in depth in order to use as an alternative to the conventional pellet fuel method. Especially, there are some questions on meeting the CANDU requirements in thermal and mechanical terms as well as density of fuel. Wherever possible, this report used representative costs of currently available technologies as the bases for cost estimation. It should also be noted that the conceptual design and cost information contained in this report was extracted from the public domain and general open literature. Later studies have to focus on other important areas of concern such as safety, security, safeguards, process optimization etc. 7 figs., 6 tabs. (Author)

HTGR spent fuel storage study

International Nuclear Information System (INIS)

Burgoyne, R.M.; Holder, N.D.

1979-04-01

This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

Modeling of constituent redistribution in U-Pu-Zr metallic fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)]. E-mail: yskim@anl.gov; Hayes, S.L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Yacout, A.M. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2006-12-01

A computer model was developed to analyze constituent redistribution in U-Pu-Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

Conceptual design study on advanced aqueous reprocessing system for fast reactor fuel cycle

International Nuclear Information System (INIS)

Takata, Takeshi; Koma, Yoshikazu; Sato, Koji; Kamiya, Masayoshi; Shibata, Atsuhiro; Nomura, Kazunori; Ogino, Hideki; Koyama, Tomozo; Aose, Shin-ichi

2003-01-01

As a feasibility study on commercialized fast reactor cycle system, a conceptual design study is being progressed for the aqueous and pyrochemical processes from the viewpoint of economical competitiveness, efficient utilization of resources, decreasing environmental impact and proliferation resistance in Japan Nuclear Cycle Development Institute (JNC). In order to meet above-mentioned requirements, the survey on a range of reprocessing technologies and the evaluation of conceptual plant designs against targets for the future fast reactor cycle system have been implemented as the fist phase of the feasibility study. For an aqueous reprocessing process, modification of the conventional PUREX process (a solvent extraction process with purification of U/Pu, with nor recovery of minor actinides (MA)) and investigation of alternatives for the PUREX process has been carried out and design study of advanced aqueous reprocessing system and its alternatives has been conducted. The conceptual design of the advanced aqueous reprocessing system has been updated and evaluated by the latest R and D results of the key technologies such as crystallization, single-cycle extraction, centrifugal contactors, recovery of Am/Cm and waste processing. In this paper, the outline of the design study and the current status of development for advanced aqueous reprocessing system, NEXT process, are mentioned. (author)

Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

International Nuclear Information System (INIS)

Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

2010-01-01

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

International Nuclear Information System (INIS)

Doshi, P.K.; DeMario, E.E.; Knott, R.P.

1993-01-01

Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

Potential of thorium-based fuel cycle for PWR core to reduce plutonium and long-term toxicity

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

The cross section libraries and calculation methods of the participants were inter-compared through the first stage benchmark calculation. The multiplication factor of unit cell benchmark are in good agreement, but there is significant discrepancies of 2.3 to 3.5 %k at BOC and at EOC between the calculated infinite multiplication factors of each participants for the assembly benchmark. Our results with HELIOS show a reasonable agreement with the others except the MTC value at EOC. To verify the potential of the thorium-based fuel to consume the plutonium and to reduce the radioactivity from the spent fuel, the conceptual core with ThO{sub 2}-PuO{sub 2} or MOX fuel were constructed. The composition and quantity of plutonium isotopes and the radioactivity level of spent fuel for conceptual cores were analyzed, and the neutronic characteristics of conceptual cores were also calculated. The nuclear characteristics for ThO{sub 2}-PuO{sub 2} thorium fueled core was similar to MOX fueled core, mainly due to the same seed fuel material, plutonium. For the capability of plutonium consumption, ThO{sub 2}-PuO{sub 2} thorium fuel can consume plutonium 2.1-2.4 times MOX fuel. The fraction of fissile plutonium in the spent ThO{sub 2}-PuO{sub 2} thorium fuel is more favorable in view of plutonium consumption and non-proliferation than MOX fuel. The radioactivity of spent ThO{sub 2}-PuO{sub 2} thorium and MOX fuel batches were calculated. Since plutonium isotopes are dominant for the long-term radioactivity, ThO{sub 2}-PuO{sub 2} thorium has almost the same level of radioactivity as in MOX fuel for a long-term perspective. (author). 22 figs., 11 tabs.

Fueling moving ring field-reversed mirror reactor plasmas

International Nuclear Information System (INIS)

Felber, F.S.

1980-01-01

The concept of small fusion reactors is being studied jointly by Lawrence Livermore Laboratory General Atomic Company, and Pacific Gas and Electric Company. The objective is to investigate alternatives and then to develop a conceptual design for a small reactor that could produce useful, though not necessarily economical, energy by the late 1980s. Three methods of fueling a small moving ring field-reversed mirror are considered: injection of fuel pellets accelerated by laser ablation, injection of fuel pellets accelerated by deflagration-gun ablation, and direct injection of plasma by a deflagration gun. 13 refs

Conceptual safety design analysis of Korea advanced liquid metal reactor

International Nuclear Information System (INIS)

Suk, S. D.; Park, C. K.

1999-01-01

The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

Coupling of the four design stages in the management of nuclear fuel; Acoplamiento de las cuatro etapas de diseno en la administracion de combustible nuclear

Energy Technology Data Exchange (ETDEWEB)

Marinez R, R. L.

2016-07-01

In this work, the main characteristics of the system to solve the four stages of the nuclear fuel management are presented; the above for boiling water reactors (BWR). The novelty of the system is that a complete solution is obtained in a coupled way; the involved stages are fuel lattice design, fuel assembly design, fuel reload design and control rod pattern design. To do this, in each stage of the process some heuristics techniques are applied, and each stage has its own objective function. The used heuristic techniques are neural network and a hybrid between scatter search and path re linking for fuel lattice design; for fuel assembly design a simple local search was applied and finally, for both fuel reload and control rod pattern designs, the tabu search technique was used. The system have two loops, one external loop and one internal loop, the first one starts with fuel lattice design and concludes with control rod pattern design; on the other hand, the internal loop executes an iterative process between both fuel reload design and control rod pattern designs, to start this loop a seed fuel reload is required, which is obtained applying Haling principle. The internal loop is finished when four iterations were achieved, while the external loop is finished when two iterations were achieved, this number of iterations was fixed due to the great quantity of required computational resources. An 18- months equilibrium cycle was considered to have a reference value to compare against the obtained results with our system, this cycle have two fuel fresh batches with the same average uranium enrichment, but different gadolinia content. The above cycle achieved a 10,896 Mwd/Tu of energy and was divided into 12 burnup steps. The obtained results show the advantage to solve the complete problem in a coupled way, even though a great quantity of computational resources are used. It is necessary to note that the energy value was not achieved in all cases, only in some

Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel

International Nuclear Information System (INIS)

1978-09-01

This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

Energy Technology Data Exchange (ETDEWEB)

Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

1994-12-31

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

Recent progress in conceptual design of LMFR in China

International Nuclear Information System (INIS)

Xu Mi; Ma Dayuan; Chen Yishao

1996-01-01

The Chinese Experimental Fast Reactor (CEFR-25) with the thermal power 65MW and electric power 25MW is the first step of the FBR development in China. The aims of this project are as following: As a prototype to accumulate the experiences of design, construction and operation of a fast reactor; as an irradiation facility to develop fuels and materials with high breeding properties and burn up rate; and as a test reactor core, envisaged, to test the fast reactor core with the fuel Ac-contained. After three years preparation for the CEFR design including the development, collection and reviewing of about 50 computer codes and the decision of the main technical selections and of design boundary conditions, from 1990 to July 1992, the conceptual design of the CEFR-25 has been completed. The confirmation and some optimization of the conceptual design have been carried out from October 1992 to the end of 1993. Based on the conceptual design and related optimization, the main features and characteristics of the CEFR-25 has been given in this paper. Spent almost whole year 1994 for the input preparation the CEFR-25 technical design has been started in the early of the year 1995. The input for the technical design mainly includes: design criteria; input parameters; technical selections confirmed in conceptual design; design requirements, etc., which are briefly described in this paper. For the design requirements to the CEFR-25, of the most important is that the reactor should have passive safety properties, in other words, during any credible transient incident, for instance, ULOF-ULOHS and UTOP, etc., by the negative feedback of reactivity the reactor will enter into and be keeping at safety condition without needs of any personal interference. The residual heat will be removed away by natural convection and natural circulation. Finally, the paper gives the general descriptions about the research and development status for this reactor. (author). 5 figs, 4 tabs

Waste Handeling Building Conceptual Study

Energy Technology Data Exchange (ETDEWEB)

G.W. Rowe

2000-11-06

The objective of the ''Waste Handling Building Conceptual Study'' is to develop proposed design requirements for the repository Waste Handling System in sufficient detail to allow the surface facility design to proceed to the License Application effort if the proposed requirements are approved by DOE. Proposed requirements were developed to further refine waste handling facility performance characteristics and design constraints with an emphasis on supporting modular construction, minimizing fuel inventory, and optimizing facility maintainability and dry handling operations. To meet this objective, this study attempts to provide an alternative design to the Site Recommendation design that is flexible, simple, reliable, and can be constructed in phases. The design concept will be input to the ''Modular Design/Construction and Operation Options Report'', which will address the overall program objectives and direction, including options and issues associated with transportation, the subsurface facility, and Total System Life Cycle Cost. This study (herein) is limited to the Waste Handling System and associated fuel staging system.

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

FIREBIRD - a conceptual design of a field reversed configuration compact torus fusion reactor (CTFR)

International Nuclear Information System (INIS)

Raman, R.; Zubrin, R.M.

1987-01-01

This paper is a summary of the work carried out by the Nuclear Engineering 512 design team at the University of Washington on a conceptual design study of a Compact-Torus (Field-Reversed) Fusion Reactor Configuration (CTFR). The primary objective of the study was to develop a reactor design that strived for high engineering power density, modest recirculating power and competitive cost of electrical power. A Conceptual design was developed for a translating field-reversed configuration reactor; based on the Physics developed by Tuszewski and Lindford at LANL and by Hoffman and Milroy at MSNW. Furthermore, it also appears possible to operate a simplified form of this reactor using a pure D-D fuel cycle after an initial D-T ignition ramp to reach the advanced fuel operating regime. One optimistic reactor so designed has a length of about 35 meters, producing a net electrical power of about 375 MWe

Design, construction and characterization of a rural solar furnace; Diseno, construccion y caracterizacion de un horno solar rural

Energy Technology Data Exchange (ETDEWEB)

Hernandez Luna, Gabriela

2001-06-15

centigrade, followed of the condition using 6 liters of oil, in which a maximum temperature of 115 centigrade was obtained and in the tests with 6 liters of water as load a maximum temperature of 95 centigrade was registered. In these tests the temperature of the oil surpasses 60 centigrade during an approximate period of 7 hours and is higher than 80 centigrade by a lapse of 5 hours. The water temperature is higher than 60 centigrade for a period of more than 5 hours and during 3 hours is higher than 80 centigrade. These results are satisfactory, because they assure the accomplishment of the process of baking inside the furnace. A calculation program was elaborated to implement the theoretical model of the concentration. This program reads the data of radiation intensity in the horizontal plane supplied by the weather station of the CIE and determines the incident radiation in the furnace collector. The theoretical model of the thermal behavior was implemented in another computer program; this program reads the data of the incident radiation in the collector of the furnace and the data of room temperature and calculates the temperature of the furnace as a time function. The model overestimates the temperature reached in the furnace, nevertheless reproduces qualitatively the thermal behavior of the same. In spite of the limitations of the theoretical model, this can be of utility in achieving the reproduction of the temperature of the load experimentally registered, through an effective coefficient of heat losses and an effective coefficient of heat capacity, characteristic of each one of the conditions of the test. [Spanish] Esta tesis presenta el diseno, construccion y caracterizacion de un horno solar tipo caja para su uso en comunidades rurales en la zona intertropical. El trabajo presentado en esta tesis parte del diseno opto-geometrico de un horno solar tipo caja propuesto por un entusiasta grupo de jovenes estudiantes (Acosta et al., 1996, Vazquez et al., 1998

Conceptual design of reactor assembly of prototype fast breeder reactor

International Nuclear Information System (INIS)

Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

1996-01-01

The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

Conceptual design of a neutral-beam injection system for the TFTR

International Nuclear Information System (INIS)

Ehlers, K.W.; Berkner, K.H.; Cooper, W.S.; Hooper, E.B.; Pyle, R.V.; Stearns, J.W.

1975-11-01

The neutral-beam injection requirements for heating and fueling the next generation of fusion reactor experiments far exceed those of present devices; the neutral-beam systems needed to meet these requirements will be large and complex. A conceptual design of a TFTR tokamak injection system to produce 120 keV deuterium-ion beams with a total power of about 80 MW is given

DUPIC fuel compatibility assessment

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

2000-03-01

The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

DUPIC fuel compatibility assessment

International Nuclear Information System (INIS)

Choi, Hang Bok; Rho, G. H.; Park, J. W. and others

2000-03-01

The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

KALIMER fuel system preliminary design description

International Nuclear Information System (INIS)

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

1998-10-01

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

Challenges and approaches in planning fuel treatments across fire-excluded forested landscapes

Science.gov (United States)

B.M. Collins; S.L. Stephens; J.J. Moghaddas; J. Battles

2010-01-01

Placing fuel reduction treatments across entire landscapes such that impacts associated with high-intensity fire are lessened is a difficult goal to achieve, largely because of the immense area needing treatment. As such, fire scientists and managers have conceptually developed and are refining methodologies for strategic placement of fuel treatments that...

Experimental fusion power reactor conceptual design study. Final report. Volume II

International Nuclear Information System (INIS)

Baker, C.C.

1976-12-01

This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following sections: (1) reactor components, (2) auxiliary systems, (3) operations, (4) facility design, (5) program considerations, and (6) conclusions and recommendations

A study on manufacturing and quality control technology of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Park, H. S.; Lee, Y. W. [and others

1997-09-01

A series of experiments are performed to verify the manufacturability of DUPIC fuel and its performance by use of HANARO test reactor. Major works performed during this research period are : analysis of manufacturing process of DUPIC fuel, fabrication technology development such as development of disassembly and decladding method of spent PWR fuel, study on the OREOX process using simulated high burnup fuel, weldability of end cap weld, and development of fabrication equipment including the conceptual and detailed design of DUPIC equipment mainly for the powder preparation, pelletization and fuel element fabrication. A study on the material properties of DUPIC fuel and performance analysis method using irradiation of test fuel was also performed. (author). 91 refs., 274 tabs., 254 figs.

A study on manufacturing and quality control technology of DUPIC fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Park, H. S.; Lee, Y. W.

1997-09-01

A series of experiments are performed to verify the manufacturability of DUPIC fuel and its performance by use of HANARO test reactor. Major works performed during this research period are : analysis of manufacturing process of DUPIC fuel, fabrication technology development such as development of disassembly and decladding method of spent PWR fuel, study on the OREOX process using simulated high burnup fuel, weldability of end cap weld, and development of fabrication equipment including the conceptual and detailed design of DUPIC equipment mainly for the powder preparation, pelletization and fuel element fabrication. A study on the material properties of DUPIC fuel and performance analysis method using irradiation of test fuel was also performed. (author). 91 refs., 274 tabs., 254 figs

Accelerator driven heavy water blanket on circulating fuel

International Nuclear Information System (INIS)

Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

1997-01-01

A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

Status of the spent fuel dry storage programme for Cernavoda NPP

International Nuclear Information System (INIS)

Radu, M.

1999-01-01

The Cernavoda NPP Unit 1 (600 MWe Standard type) is in operation since December 1996. Within the framework of the R and D Radioactive Waste and Spent Fuel Management Programme, investigations, studies and research are carried out on site identification and conceptual designs for both a Spent Fuel Interim Storage Facility and a Spent Fuel Disposal Facility. The status of the work performed in the framework of this programme as well as the situation of the spent fuel resulting from the Research Institutes will be presented in the paper. (author)

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

International Nuclear Information System (INIS)

Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

1979-01-01

Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

Review of Biojet Fuel Conversion Technologies

Energy Technology Data Exchange (ETDEWEB)

Wang, Wei-Cheng [National Renewable Energy Lab. (NREL), Golden, CO (United States); Tao, Ling [National Renewable Energy Lab. (NREL), Golden, CO (United States); Markham, Jennifer [National Renewable Energy Lab. (NREL), Golden, CO (United States); Zhang, Yanan [National Renewable Energy Lab. (NREL), Golden, CO (United States); Tan, Eric [National Renewable Energy Lab. (NREL), Golden, CO (United States); Batan, Liaw [National Renewable Energy Lab. (NREL), Golden, CO (United States); Warner, Ethan [National Renewable Energy Lab. (NREL), Golden, CO (United States); Biddy, Mary [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2016-07-01

Biomass-derived jet (biojet) fuel has become a key element in the aviation industry’s strategy to reduce operating costs and environmental impacts. Researchers from the oil-refining industry, the aviation industry, government, biofuel companies, agricultural organizations, and academia are working toward developing commercially viable and sustainable processes that produce long-lasting renewable jet fuels with low production costs and low greenhouse gas emissions. Additionally, jet fuels must meet ASTM International specifications and potentially be a 100% drop-in replacement for the current petroleum jet fuel. The combustion characteristics and engine tests demonstrate the benefits of running the aviation gas turbine with biojet fuels. In this study, the current technologies for producing renewable jet fuels, categorized by alcohols-to-jet, oil-to-jet, syngas-to-jet, and sugar-to-jet pathways, are reviewed. The main challenges for each technology pathway, including feedstock availability, conceptual process design, process economics, life-cycle assessment of greenhouse gas emissions, and commercial readiness, are discussed. Although the feedstock price and availability and energy intensity of the process are significant barriers, biomass-derived jet fuel has the potential to replace a significant portion of conventional jet fuel required to meet commercial and military demand.

Definition and conceptual design of a small fusion reactor

International Nuclear Information System (INIS)

1979-04-01

The objective of this project is to evaluate various mirror fusion reactor concepts that might result in small systems for the effective production of electrical power or stored energy (e.g., nuclear and chemical fuels). The basic two-year program goal is to select a particular concept and develop the conceptual design of a pilot plant that could provide a useful output from fusion. The pilot plant would be built and operated in the late 1980s

Stationary power fuel cell commercialization status worldwide

Energy Technology Data Exchange (ETDEWEB)

Williams, M.C. [Dept. of Energy, Morgantown, WV (United States)

1996-12-31

Fuel cell technologies for stationary power are set to play a role in power generation applications worldwide. The worldwide fuel cell vision is to provide powerplants for the emerging distributed generation and on-site markets. Progress towards commercialization has occurred in all fuel cell development areas. Around 100 ONSI phosphoric acid fuel cell (PAFC) units have been sold, with significant foreign sales in Europe and Japan. Fuji has apparently overcome its PAFC decay problems. Industry-driven molten carbonate fuel cell (MCFC) programs in Japan and the U.S. are conducting megawatt (MW)-class demonstrations, which are bringing the MCFC to the verge of commercialization. Westinghouse Electric, the acknowledged world leader in tubular solid oxide fuel cell (SOFC) technology, continues to set performance records and has completed construction of a 4-MW/year manufacturing facility in the U.S. Fuel cells have also taken a major step forward with the conceptual development of ultra-high efficiency fuel cell/gas turbine plants. Many SOFC developers in Japan, Europe, and North America continue to make significant advances.

Oak Ridge TNS program: study of fueling techniques in support of TNS development

International Nuclear Information System (INIS)

Simpson, W.

1979-12-01

The objective of this study was to survey fueling techniques and determine which approaches were compatible with the TNS requirements. Specifically, the following tasks were undertaken: survey of existing fueling concepts for use in TNS, determination of available physical properties for D-T fuel pellets, performance of preliminary load analysis of selected pellet acceleration machines, preparation of conceptual designs, and recommendations for follow-up work

Proposed method for the hydraulic design of ski-jump energy dissipators in dam spillways considering the occurrence of scour holes downstream of the structure; Metodo propuesto para el diseno hidraulico de trampolines empleados como disipadores de energia en aliviaderos para presas, considerando la ocurrencia del cono de socavacion al pie del mismo

Energy Technology Data Exchange (ETDEWEB)

Pardo-Gomez, Rafael [Centro de Investigaciones Hidraulicas (Cuba)

2008-04-15

Ski-jump energy dissipators are widely used in hydraulic engineering because of their well-known effectiveness. Nevertheless, some uncertainty exits associated with the dimensions of the scour hole appearing downstream of the structure. This paper presents a new method for solving this problem. This method includes spillway stability checking as part of the design process and also stability checking of any other construction near the energy dissipation zone. [Spanish] Los disipadores de energia tipo trampolin tienen amplia utilizacion en la practica de la ingenieria hidraulica por su probada eficacia; sin embargo, su diseno esta sujeto a cierto grado de incertidumbre en cuanto a la prediccion de las dimensiones del cono de socavacion que habra de producirse aguas abajo de la estructura. En el presente trabajo se muestra un metodo novedoso, mediante el cual el autor soluciona el aspecto antes referido, toda vez que se incluye como parte del proceso de diseno la comprobacion de la estabilidad del propio aliviadero o de cualquier otra obra cercana a la zona de disipacion de energia.

The disposal of Canada's nuclear fuel waste: engineering for a disposal facility

International Nuclear Information System (INIS)

Simmons, G.R.; Baumgartner, P.

1994-01-01

This report presents some general considerations for engineering a nuclear fuel waste disposal facility, alternative disposal-vault concepts and arrangements, and a conceptual design of a used-fuel disposal centre that was used to assess the technical feasibility, costs and potential effects of disposal. The general considerations and alternative disposal-vault arrangements are presented to show that options are available to allow the design to be adapted to actual site conditions. The conceptual design for a used-fuel disposal centre includes descriptions of the two major components of the disposal facility, the Used-Fuel Packaging Plant and the disposal vault; the ancillary facilities and services needed to carry out the operations are also identified. The development of the disposal facility, its operation, its decommissioning, and the reclamation of the site are discussed. The costs, labour requirements and schedules used to assess socioeconomic effects and that may be used to assess the cost burden of waste disposal to the consumer of nuclear energy are estimated. The Canadian Nuclear Fuel Waste Management Program is funded jointly by AECL and Ontario Hydro under the auspices of the CANDU Owners Group. (author)

Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

International Nuclear Information System (INIS)

Sample, C.R.

1977-02-01

A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL

Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

Energy Technology Data Exchange (ETDEWEB)

Sample, C R [comp.

1977-02-01

A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

Conceptual Design for BOP of the Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Yoo, Tae Geun; Kim, Seong O; Kim, Eui Kwang; Seong, Seung Hwan

2010-01-01

The heavy dependence on nuclear power eventually raise the issues of an efficient utilization of uranium resources, which Korea presently imports from abroad, end of a spent fuel storage. From the viewpoint that sodium-cooled fast Reactors (SFR s ) have the potential of an enhanced safety by utilizing inherent safety characteristics, trans-uranics (TRU) reduction and resolving the spent fuel storage problems through a proliferation-resistant actinide recycling. SFR s are sure to be most promising nuclear power operation. The Korea Atomic Energy Research Institute (KAERI) has been developing SFR design technologies since 1997. And nowadays, the preliminary heat balance of the demonstration SFR is calculated. However, in order to verify design condition of the NSSS, it is necessary to set the heat balance and the conceptual design for BOP of the SFR as a part of the SFR design technique development business. Moreover, in order to confirm whether the heat balance can actually appropriate via the turbine characteristic, it is required to carry out the performance analysis of the turbine cycle. For that, the main purposes of this study are; 1) to derivate the conceptual design for BOP, 2) to analyze the performance of the turbine cycle, 3) to derivate the main consideration for BOP design

Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor; Diseno y construccion del SIPPING para combustibles del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

2003-07-01

The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

Optimization in the design and efficiency of retractable soot blowers; Optimacion del diseno y la eficiencia de sopladores de hollin retractiles

Energy Technology Data Exchange (ETDEWEB)

Diego Marin, Antonio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1988-12-31

In this article the importance of soot blowers in the subject of design, operation and maintenance are described and the effects that its inefficient functioning causes in the steam generators. The activities and the results of a project for the evaluation of the functioning of the soot blowers in the Comision Federal de Electricidad (CFE) boilers. Finally, the scope of a new project oriented towards the retractable soot blowers efficiency optimization, and to the creation of the infra-structure to substitute the import of its components. [Espanol] En este articulo se describe la importancia de los sopladores de hollin en los aspectos de diseno, operacion y mantenimiento, y los efectos que su funcionamiento deficiente produce en los generadores de vapor. Se presentan tambien las actividades y los resultados de un proyecto para evaluar el funcionamiento de los deshollinadores de las calderas de la Comision Federal de Electricidad (CFE). Finalmente, se presenta el alcance de un nuevo proyecto que se orienta a optimar la eficiencia de los sopladores de hollin retractiles y a crear la infraestructura para sustituir las importaciones de sus componentes.

Optimization in the design and efficiency of retractable soot blowers; Optimacion del diseno y la eficiencia de sopladores de hollin retractiles

Energy Technology Data Exchange (ETDEWEB)

Diego Marin, Antonio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1987-12-31

In this article the importance of soot blowers in the subject of design, operation and maintenance are described and the effects that its inefficient functioning causes in the steam generators. The activities and the results of a project for the evaluation of the functioning of the soot blowers in the Comision Federal de Electricidad (CFE) boilers. Finally, the scope of a new project oriented towards the retractable soot blowers efficiency optimization, and to the creation of the infra-structure to substitute the import of its components. [Espanol] En este articulo se describe la importancia de los sopladores de hollin en los aspectos de diseno, operacion y mantenimiento, y los efectos que su funcionamiento deficiente produce en los generadores de vapor. Se presentan tambien las actividades y los resultados de un proyecto para evaluar el funcionamiento de los deshollinadores de las calderas de la Comision Federal de Electricidad (CFE). Finalmente, se presenta el alcance de un nuevo proyecto que se orienta a optimar la eficiencia de los sopladores de hollin retractiles y a crear la infraestructura para sustituir las importaciones de sus componentes.

Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

International Nuclear Information System (INIS)

Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

1990-08-01

A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs

Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

International Nuclear Information System (INIS)

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

2005-01-01

A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

Conceptual model for marketing solar based technology to developing countries

International Nuclear Information System (INIS)

Peter, Raj; Ramaseshan, B.; Nayar, C.V.

2002-01-01

Developing countries are faced with large shortfalls of electric-power generation, shortages of usable indigenous fuels, and transportation bottlenecks for those fuels. The loss of revenue that is forced upon the industry due to power cuts and frequent interruptions of power supply is substantial. Renewable sources of energy have an important role to play in providing much needed power in the context of growing global concern about sustainable energy supplies and protecting the environment from the adverse effects of fossil fuel utilization. The purpose of the study is to identify the factors that influence the adoption of solar-based technology. An examination of the literature in the area of diffusion of technology has led to the identification of different variables. The possible relationships that may exist between these variables is depicted in the conceptual framework. This paper attempts to clarify the relationships between the variables that have been identified and the decision to adopt. (Author)

Areva new fuel designs; increased reliability, operating margins and operating efficiency; Nuevos disenos de combustible Areva: aumento de fiabilidad, margenes operativos y eficiencia del funcionamiento

Energy Technology Data Exchange (ETDEWEB)

Mollard, P.; Vollmer, N.; Curca-Tivig, F.; Cole, S.; Louf, H. P.

2015-07-01

AREVA is continuously working on the improvement of the fuel design to address immediate and future needs of the utilities. This improvement process regularly leads to incremental changes but also to breakthrough changes addressing the next needs of the market. Since a few years now, the improvements of the fuel design and licensing benefit from the improvement and upgrade in codes and methods and computational capabilities. Changes in design are sustained by these more powerful and phenomenological tools which secure and fasten the fuel design optimization and its implementation. (Author)

Conceptual design of inherently safe integral reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, J. I.; Chang, M. H.; Lee, D. J. and others

1999-03-01

The design concept of a 300 MWt inherently safe integral reactor(ISIR) for the propulsion of extra large and superhigh speed container ship was developed in this report. The scope and contents of this report are as follows : 1. The state of the art of the technology for ship-mounted reactor 2. Design requirements for ISIR 3. Fuel and core design 4. Conceptual design of fluid system 5. Conceptual design of reactor vessel assembly and primary components 6. Performance analyses and safety analyses. Installation of two ISIRs with total thermal power of 600MWt and efficiency of 21% is capable of generating shaft power of 126,000kW which is sufficient to power a container ship of 8,000TEU with 30knot cruise speed. Larger and speedier ship can be considered by installing 4 ISIRs. Even though the ISIR was developed for ship propulsion, it can be used also for a multi-purpose nuclear power plant for electricity generation, local heating, or seawater desalination by mounting on a movable floating barge. (author)

Development of dynamic simulation code for fuel cycle fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1999-02-01

A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

System Sensitivity Analysis Applied to the Conceptual Design of a Dual-Fuel Rocket SSTO

Science.gov (United States)

Olds, John R.

1994-01-01

This paper reports the results of initial efforts to apply the System Sensitivity Analysis (SSA) optimization method to the conceptual design of a single-stage-to-orbit (SSTO) launch vehicle. SSA is an efficient, calculus-based MDO technique for generating sensitivity derivatives in a highly multidisciplinary design environment. The method has been successfully applied to conceptual aircraft design and has been proven to have advantages over traditional direct optimization methods. The method is applied to the optimization of an advanced, piloted SSTO design similar to vehicles currently being analyzed by NASA as possible replacements for the Space Shuttle. Powered by a derivative of the Russian RD-701 rocket engine, the vehicle employs a combination of hydrocarbon, hydrogen, and oxygen propellants. Three primary disciplines are included in the design - propulsion, performance, and weights & sizing. A complete, converged vehicle analysis depends on the use of three standalone conceptual analysis computer codes. Efforts to minimize vehicle dry (empty) weight are reported in this paper. The problem consists of six system-level design variables and one system-level constraint. Using SSA in a 'manual' fashion to generate gradient information, six system-level iterations were performed from each of two different starting points. The results showed a good pattern of convergence for both starting points. A discussion of the advantages and disadvantages of the method, possible areas of improvement, and future work is included.

Proposed high throughput electrorefining treatment for spent N- Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1996-01-01

A high-throughput electrorefining process is being adapted to treat spent N-Reactor fuel for ultimate disposal in a geologic repository. Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the type of fragmentation necessary to provide fuel segments suitable for this process. Based on these tests, a conceptual design was produced of a plant-scale electrorefiner. In this design, the diameter of an electrode assembly is about 1.07 m (42 in.). Three of these assemblies in an electrorefiner would accommodate a 3-metric-ton batch of N-Reactor fuel that would be processed at a rate of 42 kg of uranium per hour

Program plan for research and development in support of LWR fuel recycle

International Nuclear Information System (INIS)

1975-01-01

The ERDA program that is being planned to assist industry in the commercialization of the LWR fuel cycle will involve a range of activities, including joint programs with industry, R and D to provide technology, conceptual design of fuel recycle facilities, and environmental and economic assessments. A two-part program to begin in 1976 that is a portion of the overall ERDA plan is described. Responsibility for coordination and management of the tasks described in this document has been assigned to Du Pont as prime contractor to the ERDA Savannah River Operations Office. The first part of the program consists of the conceptual design of complete recycle facilities. The second part of the program, which will proceed concurrently, consists of supporting R and D activities, economic and environmental studies, and other studies to assist in the regulatory process. The R and D program will include both near-term activities in support of the conceptual design effort, and other activities aimed at general improvements in fuel cycle technology. The conceptual design will be used to develop current cost information for a complete reprocessing complex. The design will be based initially on current technology with provision for improvements as confirmatory information and advanced technology become available from the R and D program. The conceptual design and cost estimate will be developed by the Du Pont Atomic Energy Division. The R and D program and supporting studies will be directed at uncertainties in current technology as well as toward development of improved technology. It will include such R and D as might be appropriate for ERDA to undertake in support of joint programs with industry. The Savannah River Laboratory will have responsibility for coordinating the program

Spent nuclear fuel Canister Storage Building CDR Review Committee report

International Nuclear Information System (INIS)

Dana, W.P.

1995-12-01

The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

Centralized disassembly and packaging of spent fuel in the DOE spent fuel management system

International Nuclear Information System (INIS)

Johnson, E.R.

1986-01-01

In October 1984, E.R. Johnson Associates, Inc. (JAI) initiated a study of the prospective use of a centralized facility for the disassembly and packaging of spent fuel to support the various elements of the US Dept. of Energy (DOE) spent fuel management system, including facilities for monitored retrievable storage (MRS) and repositories. It was DOE's original plan to receive spent fuel at each repository where it would be disassembled and packaged (overpacked) for disposal purposes. Subsequently, DOE considered the prospective use of MRS of spent fuel as an option for providing safe and reliable management of spent fuel. This study was designed to consider possible advantages of the use of centralized facilities for disassembly and packaging of spent fuel at whose location storage facilities could be added as required. The study was divided into three principal technical tasks that covered: (a) development of requirements and criteria for the central disassembly and packaging facility and associated systems. (2) Development of conceptual designs for the central disassembly and packaging facility and associated systems. (3) Estimation of capital and operating costs involved for all system facilities and determination of life cycle costs for various scenarios of operation - for comparison with the reference system

Fast breeder fuel cycle

International Nuclear Information System (INIS)

1978-07-01

This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Design of a hydraulic microturbine type Michell-Banki; Diseno de una microturbina hidraulica tipo Michell-Banki

Energy Technology Data Exchange (ETDEWEB)

Lopez Luiz, Norberto; Rodriguez Bautista, Jose; Verduzco Figueroa, Hilario; Gudino Lau, Jorge; Galindo Granados, Miguel angel [Universidad de Colima, Facultad de Ingenieria Electromecanica, Manzanillo, Colima (Mexico)]. E-mail: norberto_lopez@ucol.mx

2010-11-15

The aim of this paper is to present the design of a hydraulic microturbine (MTH), which may be installed in areas where supplies are not available for electric power transmission lines, but possessing hydroelectric. MTH's role is to convert hydraulic energy into mechanical work, is coupled to an electric generator, whose function is to convert the mechanical energy into electrical energy. The turbine generator set will produce more electric power the greater the hydroelectric of the area. In southern Mexico, we have great hydroelectric potential, as the case of the states of Veracruz, Chiapas and Oaxaca, which have small causes of rivers where the MTH could be installed and put into operation. [Spanish] El objetivo de este articulo es presentar el diseno de una microturbina hidraulica (MTH), la cual podra ser instalada en zonas donde no se cuente con suministro de energia electrica por medio de lineas de transmision, pero que posean aprovechamientos hidroelectricos. La funcion de la MTH es convertir la energia hidraulica en trabajo mecanico, estara acoplada a un generador electrico, cuya funcion es convertir dicha energia mecanica en energia electrica. El conjunto turbina generador producira mas potencia electrica mientras mayor sea el aprovechamiento hidroelectrico de la zona. En el sur de Mexico contamos con un gran potencial hidroelectrico, como el caso de los estados de Veracruz, Chiapas y Oaxaca, que cuentan con pequenos causes de rios donde las MTH podrian ser instaladas y puestas en operacion.

Development of fuel performance and thermal hydraulic technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

2000-03-01

Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

Design and manufacture of an atmospheric burner of biogas with rural application; Diseno y construccion de un quemador atmosferico de biogas con aplicaciones rurales

Energy Technology Data Exchange (ETDEWEB)

Nunez Nunez, Jorge; Suarez Pacheco, Jose; Novelo Navarrete, Jose H; Soto Apolinar, Efrain [Universidad de Quintana Roo, Chetumal, Quintana Roo (Mexico)

2000-07-01

In this text it's resumed the methodology that was carried out to make an atmospheric burner of biogas, as the criteria that were taken in account in order to determine the main parameters of it. It introduces a synthesis of the stages of design and manufacture of the device. The utility of this type of burner increase the efficiency of the oxidation of the biogas compared with the use of conventional burners that aren't designed for this purpose. [Spanish] En este trabajo se resume la metodologia que se llevo a cabo para la construccion de un quemador de biogas tipo atmosferico, asi como los criterios que se tomaron para la determinacion de los parametros principales del mismo. Se presenta una sintesis de las etapas de diseno y manufactura del dispositivo. El uso de este tipo de quemadores aumenta la eficiencia de la oxidacion del biogas en comparacion con el uso de quemadores convencionales que no estan disenados para quemar biogas.

Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

International Nuclear Information System (INIS)

Gehin, J.C.; Worley, B.A.; Renier, J.P.; Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M.

1995-08-01

This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates

Greenfield Alternative Study LEU-Mo Fuel Fabrication Facility

Energy Technology Data Exchange (ETDEWEB)

Washington Division of URS

2008-07-01

This report provides the initial “first look” of the design of the Greenfield Alternative of the Fuel Fabrication Capability (FFC); a facility to be built at a Greenfield DOE National Laboratory site. The FFC is designed to fabricate LEU-Mo monolithic fuel for the 5 US High Performance Research Reactors (HPRRs). This report provides a pre-conceptual design of the site, facility, process and equipment systems of the FFC; along with a preliminary hazards evaluation, risk assessment as well as the ROM cost and schedule estimate.

Status of solid polymer electrolyte fuel cell technology and potential for transportation applications

Science.gov (United States)

McElroy, J. F.; Nuttall, L. J.

The solid polymer electrolyte (SPE) fuel cell represents the first fuel cell technology known to be used operationally. Current activities are mainly related to the development of a space regenerative fuel cell system for energy storage on board space stations, or other large orbiting vehicles and platforms. During 1981, a study was performed to determine the feasibility of using SPE fuel cells for automotive or other vehicular applications, using methanol as the fuel. The results of this study were very encouraging. Details concerning a conceptual automotive fuel cell power plant study are discussed, taking into account also a layout of major components for compact passenger car installation.

Concept and development status of fast breeder reactor fuels in the FaCT project

International Nuclear Information System (INIS)

Maeda, S.; Suzuki, M.; Kaito, T.; Tanaka, K.; Abe, T.

2013-01-01

The fuel development and the conceptual design study have been progressed in the first phase of the FaCT project in Japan. Significant outcomes of key technologies related to fuel design, fuel properties, core materials, fuel fabrication have been provided. The prospects of these technologies have been identified. After the Fukushima accident, the research and development for reducing the amount and toxic level of radioactive wastes will be promoted more than before. These outcomes will be reflected on the future development

Waste transmutation with minimal fuel cycle long-term risk

Energy Technology Data Exchange (ETDEWEB)

Slessarev, I.; Salvatores, M.; Uematsu, M. [Direction des Reacteurs Nucleaires, Cadarache (France)

1995-10-01

Hybrid systems (source-driven subcritical reactors), are investigated at CEA, mainly from a conceptual point of view, in order to assess their potential to transmute radioactive wastes (mainly long-lived fission products, LLFP) and their potential to insure a minimal long-term radiological risk related both to the fuel inventory inside the system and to the full fuel cycle (mass flows, reprocessing transport, waste disposal). The physics of these systems has been explored and work is in progress both in the field of basic data and INC code validation, in the frame of international collaborations and in the field of conceptual design studies. The most interesting feature of subcritical source-driven system is related to the possibility to obtain an {open_quotes}excess{close_quotes} of neutrons per fission, which can be used to reduce the long-term radiological risk. A specific example will be discussed here.

Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design

International Nuclear Information System (INIS)

Wendel, M.W.; Chen, N.C.J.; Yoder, G.L.

1994-01-01

The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations

Status report on the conceptual design of a commercial tokamak hybrid reactor (CTHR)

International Nuclear Information System (INIS)

1979-09-01

A preliminary conceptual design is presented for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) plants. The study has been made in sufficient depth to indicate no insurmountable technical problems exist and has provided a basis for valid cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources

Is Criminal Behavior a Central Component of Psychopathy? Conceptual Directions for Resolving the Debate

Science.gov (United States)

Skeem, Jennifer L.; Cooke, David J.

2010-01-01

The development of the Psychopathy Checklist-Revised (PCL-R; R. D. Hare, 2003) has fueled intense clinical interest in the construct of psychopathy. Unfortunately, a side effect of this interest has been conceptual confusion and, in particular, the conflating of measures with constructs. Indeed, the field is in danger of equating the PCL-R with…

Una herramienta para la selección automatizada de aceros en el contexto // A tool for the automated selection of steels in the Mechanical Engineering´s context

Directory of Open Access Journals (Sweden)

L. Dumitrescu

2010-01-01

Full Text Available ResumenEl diseno de Ingenieria de un producto o componente constituye una actividad dificil, compleja ymultidisciplinaria, enfocada a la resolucion de problemas. En el presente trabajo se muestra eldesarrollo de una herramienta automatizada para la seleccion de los aceros mas utilizados en laconstruccion de maquinarias. La herramienta constituye una ayuda para la seleccion de losmateriales desde la etapa conceptual del proceso de diseno, donde se identifican las diferentescategorias de materiales a utilizar. La herramienta comprende las caracteristicas y propiedades masrelevantes de los aceros de Ingenieria disponibles en seis normas internacionales: Japonesa (JIS,Alemana (DIN, Vbn, DIN-Vbn, Rusa (GOST, Americana (AISI, SAE, AISI-SAE, ASTM, Inglesa (BS, EN,GB, Francesa (AFNOR y la Norma Cubana (NC vigente.Palabras claves: materiales, automatizada, aceros, ingenieria.____________________________________________________________AbstractThe product or component design constitutes a difficult, complex and multidisciplinary activity,focused to the resolution of problems. Presently work show the development of an automated toolfor the selection of the more used steels in the construction of machineries. The tool constitute ahelp for the selection of the materials from the conceptual stage of the design process, where thedifferent categories of materials are identified to use. The tool analyze the characteristics and morecommon properties of the available steels in six international standards: JIS (Japan, DIN, Vbn, DINVbn(Germany, GOST (Russia, AISI, SAE, AISI-SAE, ASTM (USA, BS, IN, GB (England, AFNOR(France and NC (Cuba.Key words: materials, automated, steels, engineering.

Conceptual design tool development for a Pb-Bi cooled reactor

International Nuclear Information System (INIS)

Lee, K. G.; Chang, S. H.; No, H. C.; Chunm, M. H.

2000-01-01

Conceptual design is generally ill-structured and mysterious problem solving. This leads the experienced experts to be still responsible for the most of synthesis and analysis task, which are not amenable to logical formulations in design problems. Especially because a novel reactor such as a Pb-Bi cooled reactor is going on a conceptual design stage, it will be very meaningful to develop the conceptual design tool. This tool consists of system design module with artificial intelligence, scaling module, and validation module. System design decides the optimal structure and the layout of a Pb-Bi cooled reactor, using design synthesis part and design analysis part. The designed system is scaled to be optimal with desired power level, and then the design basis accidents (Dbase) are analyzed in validation module. Design synthesis part contains the specific data for reactor components and the general data for a Pb-Bi cooled reactor. Design analysis part contains several design constraints for formulation and solution of a design problem. In addition, designer's intention may be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool is applied to a Pb-Bi cooled reactor with 125 M Wth of power level. The Pb-Bi cooled reactor is a novel reactor concept in which the fission-generated heat is transferred from the primary coolant to the secondary coolant through a reactor vessel wall of a novel design. The Pb-Bi cooled reactor is to deliver 125 M Wth per module for 15 effective full power years without any on-site fuel handling. The conceptual design tool investigated the feasibility of a Pb-Bi cooled reactor. Application of the conceptual design tool will be, in detail, presented in the full paper. (author)

Determination of prerequisites for the estimation of transportation cost of spent fuels

International Nuclear Information System (INIS)

Choi, Heui Joo; Lee, Jong Youl; Kim, Seong Ki; Cha, Jeong Hoon; Choi, Jong Won

2007-10-01

The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The scope of this report is limited to the cost for the spent fuel transportation. KAERI is developing a cost estimation method for the spent fuel transportation through a joint study with the French AREVA TN. Several prerequisites should be fixed in order to estimate the cost for the spent fuel transportation properly. In this report we produced them considering the Korean current status on the management of spent fuels. The representative characteristics of a spent fuel generated from the six nuclear reactors at the YG site were determined. Total 7,200 tons of spent fuels are projected with the lifespan of 60 years. As the transportation mode, sea transportation and road transportation is recommended considering the location of the YG site and the hypothetical Centralized Interim Storage Facility (CISF) and Final Repository (FR). The sea route and transportation time were analyzed by using a sea distance analysis program which the NORI (National Oceanographic Research Institute) supplies on a web. Based on the results of the analysis, the shipping rates were determined. The regulations related to the spent fuel transportation were reviewed. The characteristics of the transportation vessel and a trailer were suggested. The handling and transportation systems at the YG site, Centralized Interim Storage Facility, and the Final Repository were described in detail for the purpose of the cost estimation of the spent fuel transportation. From the detail description the major components of the transportation system were determined for the conceptual design. It is believed that the conceptual design of the transportation system developed in this report will be used for the analysis of transportation logistics and the cost estimation of spent fuels

Development of dynamic simulation code for fuel cycle of fusion reactor

International Nuclear Information System (INIS)

Aoki, Isao; Seki, Yasushi; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1999-02-01

A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

Grid-connected integrated community energy system. Phase II, Stage 1, final report. Conceptual design, demand and fuel projections and cost analysis

Energy Technology Data Exchange (ETDEWEB)

1978-03-08

The Phase I Report, Grid ICES, presented the broad alternatives and implications for development of an energy system satisfying thermal demand with the co-generation of electric power, all predicated on the use of solid fuels. Participants of the system are the University of Minnesota, operator and primary thermal user, and Northern States Power Company, primary electrical user; with St. Mary's Hospital, Fairview Hospital, and Augsburg College as Add-on Customers for the thermal service (Option I). Included for consideration are the Options of (II) solid waste disposal by the Pyrolysis Method, with heat recovery, and (III) conversion of a portion of the thermal system from steam to hot water distribution to increase co-generation capability and as a demonstration system for future expansion. This report presents the conceptual design of the energy system and each Option, with the economic implications identified so that selection of the final system can be made. Draft outline of the Environmental Assessment for the project is submitted as a separate report.

Conceptual design of a quasi-homogeneous pressurized heavy water reactor to be operated in the closed Th-U233 fuel cycle

International Nuclear Information System (INIS)

1979-06-01

This paper deals with the heavy water reactor, which, from the neutron economy point of view, offers advantages over the light water reactor. Its capability to be fuelled with natural uranium has also been considered a desirable nuclear option by various countries with sufficient domestic uranium resources not wishing to be dependent on the import of enrichment and other fuel cycle services which, in addition, would draw on the foreign exchange reserves. Pressurized heavy water reactors have been designed and built according to two somewhat different versions. While the Canadian CANDU-PHWR concept uses pressure tubes in a nearly unpressurized moderator tank (calandria), the German development line takes advantage of the established and well proven LWR technology, and, thus, uses a pressure vessel design where coolant channels and the surrounding moderator are held at equal pressure. This pressure vessel type heavy water reactor which has been built on a commercial demonstration plant level at ATUCHA in Argentina is described in a companion paper where also a conceptual design for a 685 MWsub(e) PHWR is discussed

Encapsulation of spent nuclear fuel in ceramic materials

International Nuclear Information System (INIS)

Forberg, S.; Westermark, T.

1983-03-01

The international situation with regard to deposition of spent nuclear fuel is surveyed, with emphasis on encapsulation in ceramic materials. The feasibility and advantages of ceramic containers, thermodynamic stable in groundwater, are discussed as well as the possibility to ensure that stability for longevity by engineered measures. The design prerequisite are summarized and suggestions are made for a conceptual design, comprising rutile containers with stacks of coiled fuel pins. A novel technique is suggested for the homogeneous sealing of rutile containers at low temperatures. acceptable also for the fuel pin package. Key points are given for research, demonstration and verifications of the design foundations and for future improvements. Of which a few ideas are exemplified. (author)

Conceptual design of a large Spectral Shift Controlled Reactor

International Nuclear Information System (INIS)

Matzie, R.A.; Menzel, G.P.

1979-08-01

Within the framework of the Nonproliferation Alternative Systems Assessment Program (NASAP), the US Department of Energy (DOE) has sponsored the development of a conceptual design of a large Spectral Shift Controlled Reactor (SSCR). This report describes the results of the development program and assesses the performance of the conceptual SSCR on the basis of fuel resource utilization and total power costs. The point of departure of the design study was a 1270 MW(e) PWR using Combustion Engineering's System 80/sup TM/ reactor and Stone and Webster's Reference Plant Design. The initial phase of the study consisted of establishing an optimal core design for both the once-through uranium cycle and the denatured U-235/thorium cycle with uranium recycle. The performance of the SSCR was then also assessed for the denatured U-233/thorium cycle with uranium recycle and for the plutonium/thorium cycle with plutonium recycle. After the optimal core design was established, the design of the NSSS and balance of plant was developed

Conceptual design of a large Spectral Shift Controlled Reactor

International Nuclear Information System (INIS)

Matzie, R.A.; Menzel, G.P.

1979-08-01

Within the framework of the Nonproliferation Alternative Systems Assessment Program (NASAP), the US Department of Energy (DOE) has sponsored the development of a conceptual design of a large Spectral Shift Controlled Reactor (SSCR). The results are presented of the development program, and the performance of the conceptual SSCR is assessed on the basis of fuel resource utilization and total power costs. The point of departure of the design study was a 1270 MW(e) PWR using Combustion Engineering's System 80 reactor and Stone and Webster's Reference Plant Design. The initial phase of the study consisted of establishing an optimal core design for both the once-through uranium cycle and the denatured U-235/thorium cycle with uranium recycle. The performance of the SSCR was then also assessed for the denatured U-233/thorium cycle with uranium recycle and for the plutonium/thorium cycle with plutonium recycle. After the optimal core design was established, the design of the NSSS and balance of plant was developed

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

The picture superiority effect in conceptual implicit memory: a conceptual distinctiveness hypothesis.

Science.gov (United States)

Hamilton, Maryellen; Geraci, Lisa

2006-01-01

According to leading theories, the picture superiority effect is driven by conceptual processing, yet this effect has been difficult to obtain using conceptual implicit memory tests. We hypothesized that the picture superiority effect results from conceptual processing of a picture's distinctive features rather than a picture's semantic features. To test this hypothesis, we used 2 conceptual implicit general knowledge tests; one cued conceptually distinctive features (e.g., "What animal has large eyes?") and the other cued semantic features (e.g., "What animal is the figurehead of Tootsie Roll?"). Results showed a picture superiority effect only on the conceptual test using distinctive cues, supporting our hypothesis that this effect is mediated by conceptual processing of a picture's distinctive features.

Fusion fuel cycle solid radioactive wastes

International Nuclear Information System (INIS)

Gore, B.F.; Kaser, J.D.; Kabele, T.J.

1978-06-01

Eight conceptual deuterium-tritium fueled fusion power plant designs have been analyzed to identify waste sources, materials and quantities. All plant designs include the entire D-T fuel cycle within each plant. Wastes identified include radiation-damaged structural, moderating, and fertile materials; getter materials for removing corrosion products and other impurities from coolants; absorbents for removing tritium from ventilation air; getter materials for tritium recovery from fertile materials; vacuum pump oil and mercury sludge; failed equipment; decontamination wastes; and laundry waste. Radioactivity in these materials results primarily from neutron activation and from tritium contamination. For the designs analyzed annual radwaste volume was estimated to be 150 to 600 m 3 /GWe. This may be compared to 500 to 1300 m 3 /GWe estimated for the LMFBR fuel cycle. Major waste sources are replaced reactor structures and decontamination waste

ANTQ evolutionary algorithm applied to nuclear fuel reload problem

International Nuclear Information System (INIS)

Machado, Liana; Schirru, Roberto

2000-01-01

Nuclear fuel reload optimization is a NP-complete combinatorial optimization problem where the aim is to find fuel rods' configuration that maximizes burnup or minimizes the power peak factor. For decades this problem was solved exclusively using an expert's knowledge. From the eighties, however, there have been efforts to automatize fuel reload. The first relevant effort used Simulated Annealing, but more recent publications show Genetic Algorithm's (GA) efficiency on this problem's solution. Following this direction, our aim is to optimize nuclear fuel reload using Ant-Q, a reinforcement learning algorithm based on the Cellular Computing paradigm. Ant-Q's results on the Travelling Salesmen Problem, which is conceptually similar to fuel reload, are better than the GA's ones. Ant-Q was tested on fuel reload by the simulation of the first cycle in-out reload of Bibils, a 193 fuel element PWR. Comparing An-Q's result with the GA's ones, it can b seen that even without a local heuristics, the former evolutionary algorithm can be used to solve the nuclear fuel reload problem. (author)

Hot conditioning equipment conceptual design report

International Nuclear Information System (INIS)

Bradshaw, F.W.

1996-01-01

This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage

Hot conditioning equipment conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Bradshaw, F.W., Westinghouse Hanford

1996-08-06

This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage.

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - IV: DUPIC Fuel Cycle Cost

International Nuclear Information System (INIS)

Ko, Won Il; Choi, Hangbok; Yang, Myung Seung

2001-01-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW.h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW.h. Considering the uncertainty (0.40 to 0.44 mills/kW.h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ∼20% and reduce the spent fuel arising by ∼65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option

Safeguards and security aspects of a potential Canadian used-fuel disposal facility

International Nuclear Information System (INIS)

Smith, R.M.; Wuschke, D.; Baumgartner, P.

1994-09-01

Large quantities of highly radioactive used fuel have been produced by Canadian nuclear generating stations. Conceptual design and development is under way to assess a means of disposing of this used fuel within a vault located 500 to 1000 m deep in plutonic rock in the Canadian Shield. In parallel with this work, the safeguards and physical security measures that will be required for this used fuel during transportation, packaging, and containment in a disposal vault are being studied in Canada, in several other countries that have similar requirements and by the International Atomic Energy Agency. Canadian commitments and regulations applicable to used-fuel transportation and disposal are described. The experience gained from applying safeguards and physical security measures at similar facilities is considered together with the availability of equipment that might be used in applying these measures. Possible safeguards and physical security measures are outlined and considered. These measures are based on the conceptual design studies for a reference Used-Fuel Disposal Centre and associated transportation systems undertaken by Atomic Energy of Canada Limited and Ontario Hydro. These studies show that effective and practical safeguards, which meet present IAEA objectives, can be applied to the used fuel in transportation and at a disposal facility. They also show that physical security measures can be employed that have a high probability of preventing theft or sabotage. 27 refs., 8 figs., 3 tabs., glossary, 2 appendices

Safeguards and security aspects of a potential Canadian used-fuel disposal facility

Energy Technology Data Exchange (ETDEWEB)

Smith, R M; Wuschke, D; Baumgartner, P

1994-09-01

Large quantities of highly radioactive used fuel have been produced by Canadian nuclear generating stations. Conceptual design and development is under way to assess a means of disposing of this used fuel within a vault located 500 to 1000 m deep in plutonic rock in the Canadian Shield. In parallel with this work, the safeguards and physical security measures that will be required for this used fuel during transportation, packaging, and containment in a disposal vault are being studied in Canada, in several other countries that have similar requirements and by the International Atomic Energy Agency. Canadian commitments and regulations applicable to used-fuel transportation and disposal are described. The experience gained from applying safeguards and physical security measures at similar facilities is considered together with the availability of equipment that might be used in applying these measures. Possible safeguards and physical security measures are outlined and considered. These measures are based on the conceptual design studies for a reference Used-Fuel Disposal Centre and associated transportation systems undertaken by Atomic Energy of Canada Limited and Ontario Hydro. These studies show that effective and practical safeguards, which meet present IAEA objectives, can be applied to the used fuel in transportation and at a disposal facility. They also show that physical security measures can be employed that have a high probability of preventing theft or sabotage. 27 refs., 8 figs., 3 tabs., glossary, 2 appendices.

A Deformation Model of TRU Metal Dispersion Fuel Rod for HYPER

International Nuclear Information System (INIS)

Lee, Byoung Oon; Hwang, Woan; Park, Won S.

2002-01-01

Deformation analysis in fuel rod design is essential to assure adequate fuel performance and integrity under irradiation conditions. An in-reactor performance computer code for a dispersion fuel rod is being developed in the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRU-Zr)-Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appeared that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel. Some experimental tests including in-pile and out-pile experiments are needed for verifying the predictive capability of the DIMAC code. An in-reactor performance analysis computer code for blanket fuel is being developed at the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRUZr)- Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appears that the deformation by swelling within fuel meat is very large for both fuels, and the major deformation mechanism at cladding is creep. The swelling strain is almost constant within the fuel meat, and is assumed to be zero in the cladding made of HT9. It is estimated that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel, and the dispersion fuel performance may be limited by swelling. But the predicted volume change of the (TRU-Zr)-Zr dispersion fuel models is about 6.1% at 30 at.% burnup. The value of cladding

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

Energy Technology Data Exchange (ETDEWEB)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

Conceptual evaluation of type B(U) casks for the nuclear power plants of Argentina

International Nuclear Information System (INIS)

Florido, P.C.; Isnardi, E.R.

1993-01-01

In Argentina two different nuclear power plants are in operation, Atucha I (PHWR-Siemens) and Embalse (PHWR-CANDU). Thus two very different fuel elements could be potentially transported. In order to optimize the research and development needed for the design and construction of the cask, the cost-benefit and flexibility of the engineering solutions are studied, for the two fuel elements. Different casks, for both types of existing fuel elements (Atucha I and Embalse), for different burnup-levels (regarding the advanced fuel cycle available), decay times, distances, and transported weight were studied. Three materials for shielding were used: uranium lead and steel. Only transport by road was considered, due to the reduced availability of the train. In this stage (conceptual design) small, easy and fast computer programs should be used. The principal issues that have to be fixed are shielding properties, thermal effects and mechanical behavior. As a result of the evaluation, different options for the casks were founded, as well as the importance of different parameters and the effect of two different designs of fuel elements. (author)

Analysis of Subjective Conceptualizations Towards Collective Conceptual Modelling

DEFF Research Database (Denmark)

Glückstad, Fumiko Kano; Herlau, Tue; Schmidt, Mikkel Nørgaard

2013-01-01

This work is conducted as a preliminary study for a project where individuals' conceptualizations of domain knowledge will thoroughly be analyzed across 150 subjects from 6 countries. The project aims at investigating how humans' conceptualizations differ according to different types of mother la...

Safeguardability of advanced spent fuel conditioning process

Energy Technology Data Exchange (ETDEWEB)

Li, T. K. (Tien K.); Lee, S. Y. (Sang Yoon); Burr, Tom; Russo, P. A. (Phyllis A.); Menlove, Howard O.; Kim, H. D.; Ko, W. I. (Won Il); Park, S. W.; Park, H. S.

2004-01-01

The Advanced Spent Fuel Conditioning Process (ACP) is an electro-metallurgical treatment technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology since 1977 for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. By using of this technology, a significant reduction of the volume and heat load of spent fuel is expected, which would lighten the burden of final disposal in terms of disposal size, safety and economics. In the framework of collaboration agreement to develop the safeguards system for the ACP, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and the KAERI since 2002. In this study, the safeguardability of the ACP technology was examined for the pilot-scale facility. The process and material flows were conceptually designed, and the uncertainties in material accounting were estimated with international target values.

A Conceptual Space Logic

DEFF Research Database (Denmark)

Nilsson, Jørgen Fischer

1999-01-01

Conceptual spaces have been proposed as topological or geometric means for establishing conceptual structures and models. This paper, after briey reviewing conceptual spaces, focusses on the relationship between conceptual spaces and logical concept languages with operations for combining concepts...... to form concepts. Speci cally is introduced an algebraic concept logic, for which conceptual spaces are installed as semantic domain as replacement for, or enrichment of, the traditional....

Investigation into rationalization of low decontamination pellet fuel fabrication plant configuration

International Nuclear Information System (INIS)

Maekawa, Kazuhiko; Yoshimura, Tadahiro; Hoshino, Yasushi; Munekata, Hideki; Tamaki, Yoshihisa

2005-02-01

In feasibility studies on commercialized FBR cycle system, a comprehensive system investigation and properties evaluation for candidate FBR cycle systems has been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation into rationalization of low decontamination pellet fuel fabrication plant configuration was carried out. Until last fiscal year, conceptual design studies of the fuel fabrication plant in 200t-HM/y scale were conducted, and system properties data concerning economics and environmental burden reduction of fuel fabrication plant was acquired. In addition to this, 50t-HM/y scale plant was also schematically studied. In this fiscal year, a rationalization study on conceptual design of 50t-HM/y scale plant was conducted with main aim of economic improvement, and the 200t-HM/y scale plant design was revised based on the recent R and D progress. The system properties data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In both case of the 50t-HM/y and 200t-HM/y scale plant, it was suggested that the equipment costs were reduced in several percentages because of reduction of maintenance equipments and cut in line number at the pellet fabrication process although granulation process fro denitration converted powder and O/M control process for pellets were added. System properties data for comparative evaluation of candidate fuel fabrication systems was also prepared. (author)

Fuel Cell Handbook, Fifth Edition

Energy Technology Data Exchange (ETDEWEB)

Energy and Environmental Solutions

2000-10-31

with proven cell designs, focusing instead on advancing the system surrounding the fuel cell to lower life cycle costs. Section 9, Fuel Cell Systems, has been significantly revised to characterize near-term and next-generation fuel cell power plant systems at a conceptual level of detail. Section 10 provides examples of practical fuel cell system calculations. A list of fuel cell URLs is included in the Appendix. A new index assists the reader in locating specific information quickly.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

Energy Technology Data Exchange (ETDEWEB)

Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

International Nuclear Information System (INIS)

Montierth, Leland M.

2016-01-01

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

Design study of a fusion-driven tokamak hybrid reactor for fissile fuel production. Final report

International Nuclear Information System (INIS)

Rose, R.P.

1979-05-01

This study evaluated conceptual approaches for a tokamak fusion-driven fuel producing reactor. The conceptual design of this hybrid reactor was based on using projected state-of-the-art technology for the late 1980s. This reactor would be a demonstration plant and, therefore, first-of-a-kind considerations have been included. The conceptual definitions of two alternatives for the fusion driver were evaluated. A Two-Component Tokamak (TCT) concept, based on the TFTR plasma physics parameters, was compared to a Beam-Driven Thermonuclear (BDTN) concept, based on the USSR T-20 plasma physics parameters

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1991-01-01

In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1990-11-01

The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

Conceptual design study FY 1981: synfuels from fusion - using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

International Nuclear Information System (INIS)

Krikorian, O.H.

1982-01-01

This report represents the second year's effort of a scoping and conceptual design study being conducted for the express purpose of evaluating the engineering potential of producing hydrogen by thermochemical cycles using a tandem mirror fusion driver. The hydrogen thus produced may then be used as a feedstock to produce fuels such as methane, methanol, or gasoline. The main objective of this second year's study has been to obtain some approximate cost figures for hydrogen production through a conceptual design study

Conceptualizing Programme Evaluation

Science.gov (United States)

Hassan, Salochana

2013-01-01

The main thrust of this paper deals with the conceptualization of theory-driven evaluation pertaining to a tutor training programme. Conceptualization of evaluation, in this case, is an integration between a conceptualization model as well as a theoretical framework in the form of activity theory. Existing examples of frameworks of programme…

The Conceptual Framework of Thematic Mapping in Case Conceptualization.

Science.gov (United States)

Ridley, Charles R; Jeffrey, Christina E

2017-04-01

This article, the 3rd in a series of 5, introduces the conceptual framework for thematic mapping, a novel approach to case conceptualization. The framework is transtheoretical in that it is not constrained by the tenets or concepts of any one therapeutic orientation and transdiagnostic in that it conceptualizes clients outside the constraints of diagnostic criteria. Thematic mapping comprises 4 components: a definition, foundational principles, defining features, and core concepts. These components of the framework, deemed building blocks, are explained in this article. Like the foundation of any structure, the heuristic value of the method requires that the building blocks have integrity, coherence, and sound anchoring. We assert that the conceptual framework provides a solid foundation, making thematic mapping a potential asset in mental health treatment. © 2017 Wiley Periodicals, Inc.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Graphs for the best hydraulic design of dropping irrigation systems of low load; Graficas para el diseno hidraulico optimo de sistemas de riego por goteo de baja carga

Energy Technology Data Exchange (ETDEWEB)

Vazquez Fernandez, E; Lopez Telles, P [Instituto de ingenieria, Universidad Nacional Autonoma de Mexico, Mexico, D.F. (Mexico)

2003-04-15

In this article are shown the graphics for the best design of dropping irrigation systems of low head with 17 mm of external diameter hose and O-Tiff type insertion dropper of low outlay. The design consists in getting the minimum number of laterals once the space between these and the droppers is established. The graphics were obtained by using a model for open networks of pipes adjusted with the information for top lengths of the dropping lateral included on the manufacture's technical sheet with an 0.90 emission uniformity for flat ground conditions. The systems can be fixed or mobile (from 0.5 to 3.2 m between laterals) to irrigate orchards or farmings with its plants separated from 0.3 to 0.8 m with an average expenditure of 1.39 L/hr for black droppers. The load on the feed-tank (or bolster) is of 1.5 m. The irrigation areas may fluctuate between 750 up to 6 780 m{sup 2} and the costs from 1.1 to 13 pesos per m{sup 2}. [Spanish] Se presentan graficas para el diseno optimo de sistemas de riego por goteo de baja carga con mangueras de 17 mm de diametro externo y goteros de insercion tipo laberinto O-Tif de goteo pequeno. El diseno consiste en obtener el numero optimo de laterales una vez establecidas las separaciones entre estos y los emisores. Las graficas se obtuvieron con ayuda de un modelo matematico para redes abiertas de tuberias que fue calibrado a partir de la informacion de longitudes maximas del lateral de goteo incluida en la hoja tecnica del fabricante, con uniformidad de emision de 0.90 en terreno plano. Los sistemas pueden ser fijos o moviles para el riego de hortalizas o cultivos con separaciones entre plantas de 0.3 a 0.8 m y de 0.5 a 3.2 m entre laterales, con gastos promedio de 1.391/h para goteros de color negro. La carga en el tanque de alimentacion o cabezales de 1.5 m. Las areas de riego van desde 750 hasta 6780 m{sup 2} y los costos unitarios de 1.1 a 1.3 pesos por m{sup 2}.

High energy density additives for Hybrid Fuel Rockets to Improve Performance and Enhance Safety

Data.gov (United States)

National Aeronautics and Space Administration — We propose a conceptual study of prototype strained hydrocarbon molecules as high energy density additives for hybrid rocket fuels to boost the performance of these...

Comparison of two thorium fuel cycles for use in light water prebreeder/breeder reactor systems (AWBA Development Program)

International Nuclear Information System (INIS)

Merriman, F.C.; McCoy, D.F.; Boyd, W.A.; Dwyer, J.R.

1983-05-01

Light water prebreeder/breeder conceptual reactor systems have been developed which have the potential to significantly improve the fuel utilization of present generation light water reactors. The purpose of this study is to describe and compare two possible types of thorium fuel cycles for use in these light water prebreeder and breeder concepts. The two types of thorium fuel cycles basically differ in the fuel rod design used in the prebreeder cores and the uranium isotopic concentration of fuel supplied to the breeder cores

Nuclear spent fuel dry storage in the EWA reactor shaft

International Nuclear Information System (INIS)

Mieleszczenko, W.; Moldysz, A.; Hryczuk, A.; Matysiak, T.

2001-01-01

The EWA reactor was in operation from 1958 until February 1995. Then it was subjected to the decommissioning procedure. Resulting from a prolonged operation of Polish research reactors a substantial amount of nuclear spent fuel of various types, enrichment and degree of burnup have been accumulated. The technology of storage of spent nuclear fuel foresees the two stages of wet storing in a water pool (deferral period from tens to several dozens years) and dry storing (deferral period from 50 to 80 years). In our case the deferral time in the water environment is pretty significant (the oldest fuel elements have been stored in water for more than 40 years). Though the state of stored fuel elements is satisfactory, there is a real need for changing the storage conditions of spent fuel. The paper is covering the description of philosophy and conceptual design for construction of the spent fuel dry storage in the decommissioned EWA reactor shaft. (author)

Two-dimensional nucleonics calculations for a ''FIRST STEP'' conceptual ICF reactor

International Nuclear Information System (INIS)

Davidson, J.W.; Battat, M.E.; Saylor, W.W.; Pendergrass, J.H.; Dudziak, D.J.

1985-01-01

A detailed two-dimensional nucleonic analysis has been performed for the FIRST STEP conceptual ICF reactor blanket design. The reactor concept incorporated in this design is a modified wetted-wall cavity with target illumination geometry left as a design variable. The 2-m radius spherical cavity is surrounded by a blanket containing lithium and 238 U as fertile species and also as energy multipliers. The blanket is configured as 0.6-m-thick cylindrical annuli containing modified LMFBR-type fuel elements with 0.5-m-thick fuel-bearing axial end plugs. Liquid lithium surrounds the inner blanket regions and serves as the coolant for both the blanket and the first wall. The two-dimensional analysis of the blanket performance was made using the 2-D discrete-ordinates code TRISM, and benchmarked with the 3-D Monte Carlo code MCNP. Integral responses including the tritium breeding ratio (TBR), plutonium breeding ratio (PUBR), and blanket energy multiplication were calculated for axial and radial blanket regions. Spatial distributions were calculated for steady-state rates of fission, neutron heating, prompt gamma-ray heating, and fuel breeding

Status report: conceptual fuel cycle studies for the Hanford Nuclear Energy Center

International Nuclear Information System (INIS)

Merrill, E.T.; Fleischman, R.M.

1975-07-01

A summary is presented of the current status of studies to determine the logistics of onsite plutonium recycle and the timing involved in introducing the associated reprocessing and fabrication fuel cycle facilities at the Hanford Nuclear Energy Center

Speeding the transition: Designing a fuel-cell hypercar

Energy Technology Data Exchange (ETDEWEB)

Williams, B.D.; Moore, T.C.; Lovins, A.B. [Rocky Mountain Inst., Snowmass, CO (United States). Hypercar Center

1997-12-31

A rapid transformation now underway in automotive technology could accelerate the transition to transportation powered by fuel cells. Ultralight, advanced-composite, low-drag, hybrid-electric hypercars--using combustion engines--could be three- to fourfold more efficient and one or two orders of magnitude cleaner than today`s cars, yet equally safe, sporty, desirable, and (probably) affordable. Further, important manufacturing advantages--including low tooling and equipment costs, greater mechanical simplicity, autobody parts consolidation, shorter product cycles, and reduced assembly effort and space--permit a free-market commercialization strategy. This paper discusses a conceptual hypercar powered by a proton-exchange-membrane fuel cell (PEMFC). It outlines the implications of platform physics and component selection for the vehicle`s mass budget and performance. The high fuel-to-traction conversion efficiency of the hypercar platform could help automakers overcome the Achilles` heel of hydrogen-powered vehicles: onboard storage. Moreover, because hypercars would require significantly less tractive power, and even less fuel-cell power, they could adopt fuel cells earlier, before fuel cells` specific cost, mass, and volume have fully matured. In the meantime, commercialization in buildings can help prepare fuel cells for hypercars. The promising performance of hydrogen-fueled PEMFC hypercars suggests important opportunities in infrastructure development for direct-hydrogen vehicles.

US--EC fuel cycle study: Background document to the approach and issues

International Nuclear Information System (INIS)

1992-11-01

In February 1991, DOE and the Commission of the European Communities (EC), signed a joint statement regarding the external costs of fuel cycles. This 18-month agreement committed their respective organizations to ''develop a comparative analytical methodology and to develop the best range of estimates of external costs from secondary sources'' for eight fuel cycles and four conservation options. In our study, a fuel cycle is defined as the series of physical and chemical processes and activities that are required to generate electricity from a specific fuel or resource. This foundation phase of the study is primarily limited to developing and demonstrating methods for estimating impacts and their monetized value, what we term ''damages'' or ''benefits,'' leaving aside the extent to which such damages have been internalized. However, Appendix C provides the conceptual framework for evaluating the extent of internalization. This report is a background document to introduce the study approach and to discuss the major conceptual and practical issues entailed by the incremental damage problem. As a background document, the report seeks to communicate an overview of the study and the important methodological choices that were made to conduct the research. In successive sections of the report, the methodological tools used in the study are discussed; the ecological and health impacts are reviewed using the coal fuel cycle as a reference case; and, in the final chapter, the methods for valuing impacts are detailed

Design of a grounding network in electrical substations. Materials and formulas the most used; Diseno de redes de tierra en subestaciones electricas. Materiales y formulas mas utilizadas

Energy Technology Data Exchange (ETDEWEB)

Raull Martin, J. [Facultad de Ingenieria, UNAM, Mexico, D.F. (Mexico)

2000-09-01

The purpose of this paper is to present a summarized version of the different types of construction of grounding systems, as well as their materials and formulas, which are necessary and useful for the design and building of a grounding network for medium to high voltage facilities. A description is also made of how to determine the electrical resistivity of subsoil layers with different physical characteristics. Several illustrative examples are also solved in the paper. [Spanish] El proposito de este articulo es presentar de manera resumida los diferentes tipos de construccion de los sistemas de tierra, asi como sus materiales y formulas, los cuales son necesarios y utiles para el diseno y edificacion de una red de tierra para instalaciones de mediana y alta tension. Se describe tambien la obtencion de las diferentes resistividades que presenta los terrenos con diversas caracteristicas fisicas. Asimismo, se solucionan ejercicios ilustrativos a lo largo del articulo.

Dependence of sputtering erosion on fuel-pellet characteristics

International Nuclear Information System (INIS)

Bohachevsky, I.O.; Hafer, J.F.

1977-11-01

Conceptual designs of fusion reactors operating on the principle of inertial confinement require that the dependence of cavity-wall erosion on fuel-pellet energy yield, its mass, and representative atomic number be known. A simple approximate model of sputtering erosion is presented and explicit formulas are derived that express the total amount of eroded wall material in terms of the above three parameters

Multi-Fuel Rotary Engine for General Aviation Aircraft

Science.gov (United States)

Jones, C.; Ellis, D. R.; Meng, P. R.

1983-01-01

Design studies, conducted for NASA, of Advanced Multi-fuel General Aviation and Commuter Aircraft Rotary Stratified Charge Engines are summarized. Conceptual design studies of an advanced engine sized to provide 186/250 shaft KW/HP under cruise conditions at 7620/25,000 m/ft. altitude were performed. Relevant engine development background covering both prior and recent engine test results of the direct injected unthrottled rotary engine technology, including the capability to interchangeably operate on gasoline, diesel fuel, kerosene, or aviation jet fuel, are presented and related to growth predictions. Aircraft studies, using these resultant growth engines, define anticipated system effects of the performance and power density improvements for both single engine and twin engine airplanes. The calculated results indicate superior system performance and 30 to 35% fuel economy improvement for the Rotary-engine airplanes as compared to equivalent airframe concept designs with current baseline engines. The research and technology activities required to attain the projected engine performance levels are also discussed.

Conceptual design study of small long-life PWR based on thorium cycle fuel

International Nuclear Information System (INIS)

Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

2014-01-01

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of 233 U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation

Conceptual design of shaft seals for a nuclear waste disposal vault

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-04-01

The concept of a disposal vault in the Canadian Shield for the effective isolation of nuclear fuel wastes is being assessed as part of the Canadian Nuclear Fuel Waste Management Program. The vault would be accessed from the surface by a number of shafts, which would likely penetrate the vault environment and intersect significant rock fractures and thereby form preferential pathways for the migration of radionuclides from the disposal area to the biosphere. Golder Associates were retained to conduct a conceptual design study of sealing and backfilling the shafts. The first volume of this report reviews current shaft sinking and lining technologies, and recommends the preferred construction methods for the shafts. Factors that could affect the design of a shaft seal system are reviewed, and a conceptual shaft seal is proposed. The second volume addresses the performance assessment of a shaft seal system. While there are no specific performance criteria against which to compare the anticipated containment characteristics of the shaft seal system proposed, the methodology developed for the performance assessment of the reference design should enable the design to be modified to meet performance criteria as they are developed. The report estimates that it will cost $133.7 million in 1986 Canadian dollars to seal three reference shafts, including $18 million for labour and equipment, $103.4 million for backfill and sealing materials, $9.5 million for project indirect costs, and $2.8 million project management. (author). 53 refs., 36 tabs., 43 figs.

Conceptual design of shaft seals for a nuclear waste disposal vault

International Nuclear Information System (INIS)

1993-04-01

The concept of a disposal vault in the Canadian Shield for the effective isolation of nuclear fuel wastes is being assessed as part of the Canadian Nuclear Fuel Waste Management Program. The vault would be accessed from the surface by a number of shafts, which would likely penetrate the vault environment and intersect significant rock fractures and thereby form preferential pathways for the migration of radionuclides from the disposal area to the biosphere. Golder Associates were retained to conduct a conceptual design study of sealing and backfilling the shafts. The first volume of this report reviews current shaft sinking and lining technologies, and recommends the preferred construction methods for the shafts. Factors that could affect the design of a shaft seal system are reviewed, and a conceptual shaft seal is proposed. The second volume addresses the performance assessment of a shaft seal system. While there are no specific performance criteria against which to compare the anticipated containment characteristics of the shaft seal system proposed, the methodology developed for the performance assessment of the reference design should enable the design to be modified to meet performance criteria as they are developed. The report estimates that it will cost $133.7 million in 1986 Canadian dollars to seal three reference shafts, including $18 million for labour and equipment, $103.4 million for backfill and sealing materials, $9.5 million for project indirect costs, and $2.8 million project management. (author). 53 refs., 36 tabs., 43 figs

Experimental fusion power reactor conceptual design study. Final report. Volume III

International Nuclear Information System (INIS)

Baker, C.C.

1976-12-01

This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

Interacting Conceptual Spaces

OpenAIRE

Bolt, Josef; Coecke, Bob; Genovese, Fabrizio; Lewis, Martha; Marsden, Daniel; Piedeleu, Robin

2016-01-01

We propose applying the categorical compositional scheme of [6] to conceptual space models of cognition. In order to do this we introduce the category of convex relations as a new setting for categorical compositional semantics, emphasizing the convex structure important to conceptual space applications. We show how conceptual spaces for composite types such as adjectives and verbs can be constructed. We illustrate this new model on detailed examples.

Uranium-thorium fuel cycle in a very high temperature hybrid system

International Nuclear Information System (INIS)

Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

2011-01-01

Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

Noble gas confinement for reactor fuel melting accidents

International Nuclear Information System (INIS)

Monson, P.R.

1984-01-01

In the unlikely event of a fuel melting accident, radioactive material would be released into the reactor room. This radioactive material would consist of particulate matter, iodine, tritium, and the noble gases krypton and xenon. In the case of reactors with containment domes the gases would be contained for subsequent cleanup. For reactors without contaiment the particulates and the iodine can be effectively removed with HEPA and carbon filters of current technology; however, noble gases cannot be easily removed and would be released to the atmosphere. In either case, it would be highly desirable to have a system that could be brought online to treat this contaminated air to minimize the population dose. A low temperature adsorption system has been developed at the Savannah River Laboratory to remove the airborne radioactive material from such a fuel melting accident. Over two dozen materials have been tested in extensive laboratory studies, and hydrogen mordenite and silver mordenite were found to be the most promising adsorbents. A full-scale conceptual design has also been developed. Results of the laboratory studies and the conceptual design are discussed along with plans for further development of this concept

Noble gas confinement for reactor fuel melting accidents

International Nuclear Information System (INIS)

Monson, P.R.

1985-01-01

In the unlikely event of a fuel melting accident radioactive material would be released into the reactor room. This radioactive material would consist of particulate matter, iodine, tritium, and the noble gases krypton and xenon. In the case of reactors with containment domes, the gases would be contained for subsequent cleanup. For reactors without containment the particulates and the iodine can be effectively removed with HEPA and carbon filters of current technology; however, noble gases cannot be easily removed and would be released to the atmosphere. In either case, it would be highly desirable to have a system that could be brought online to treat this contaminated air to minimize the population dose. A low temperature adsorption system has been developed at the Savannah River Laboratory to remove the airborne radioactive material from such a fuel melting accident. Over two dozen materials have been tested in extensive laboratory studies, and hydrogen mordensite and silver mordenite were found to be the most promising absorbents. A full-scale conceptual design has also been developed. Results of the laboratory studies and the conceptual design will be discussed along with plans for further development of this concept

Technology, safety, and costs of decommissioning a reference nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Schneider, K.J.; Jenkins, C.E.; Rhoads, R.E.

1977-09-01

Safety and cost information were developed for the conceptual decommissioning of a fuel reprocessing plant with characteristics similar to the Barnwell Nuclear Fuel Plant. The main process building, spent fuel receiving and storage station, liquid radioactive waste storage tank system, and a conceptual high-level waste-solidification facility were postulated to be decommissioned. The plant was conceptually decommissioned to three decommissioning states or modes; layaway, protective storage, and dismantlement. Assuming favorable work performance, the elapsed time required to perform the decommissioning work in each mode following plant shutdown was estimated to be 2.4 years for layaway, 2.7 years for protective storage, and 5.2 years for dismantlement. In addition to these times, approximately 2 years of planning and preparation are required before plant shutdown. Costs, in constant 1975 dollars, for decommissioning were estimated to be $18 million for layaway, $19 million for protective storage and $58 million for dismantlement. Maintenance and surveillance costs were estimated to be $680,000 per year after layaway and $140,000 per year after protective storage. The combination mode of protective storage followed by dismantlement deferred for 10, 30, and 100 years was estimated to cost $64 million, $67 million and $77 million, respectively, in nondiscounted total 1975 dollars. Present values of these costs give reduced costs as dismantlement is deferred. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year radiation dose commitment to the members of the public from airborne releases from normal decommissioning activities were estimated to be less than 11 man-rem

Technology, safety, and costs of decommissioning a reference nuclear fuel reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Schneider, K.J.; Jenkins, C.E.; Rhoads, R.E.

1977-09-01

Safety and cost information were developed for the conceptual decommissioning of a fuel reprocessing plant with characteristics similar to the Barnwell Nuclear Fuel Plant. The main process building, spent fuel receiving and storage station, liquid radioactive waste storage tank system, and a conceptual high-level waste-solidification facility were postulated to be decommissioned. The plant was conceptually decommissioned to three decommissioning states or modes; layaway, protective storage, and dismantlement. Assuming favorable work performance, the elapsed time required to perform the decommissioning work in each mode following plant shutdown was estimated to be 2.4 years for layaway, 2.7 years for protective storage, and 5.2 years for dismantlement. In addition to these times, approximately 2 years of planning and preparation are required before plant shutdown. Costs, in constant 1975 dollars, for decommissioning were estimated to be $18 million for layaway, $19 million for protective storage and $58 million for dismantlement. Maintenance and surveillance costs were estimated to be $680,000 per year after layaway and $140,000 per year after protective storage. The combination mode of protective storage followed by dismantlement deferred for 10, 30, and 100 years was estimated to cost $64 million, $67 million and $77 million, respectively, in nondiscounted total 1975 dollars. Present values of these costs give reduced costs as dismantlement is deferred. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year radiation dose commitment to the members of the public from airborne releases from normal decommissioning activities were estimated to be less than 11 man-rem.

Conceptual design of a lunar oxygen pilot plant Lunar Base Systems Study (LBSS) task 4.2

Science.gov (United States)

1988-01-01

The primary objective was to develop conceptual designs of two pilot plants to produce oxygen from lunar materials. A lunar pilot plant will be used to generate engineering data necessary to support an optimum design of a larger scale production plant. Lunar oxygen would be of primary value as spacecraft propellant oxidizer. In addition, lunar oxygen would be useful for servicing nonregenerative fuel cell power systems, providing requirements for life support, and to make up oxygen losses from leakage and airlock cycling. Thirteen different lunar oxygen production methods are described. Hydrogen reduction of ilmenite and extraction of solar-wind hydrogen from bulk lunar soil were selected for conceptual design studies. Trades and sensitivity analyses were performed with these models.

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Energy Technology Data Exchange (ETDEWEB)

Calabrese, R., E-mail: rolando.calabrese@enea.i [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Vettraino, F.; Artioli, C. [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Sobolev, V. [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Thetford, R. [Serco Technical and Assurance Services, 150 Harwell Business Centre, Didcot OX11 0QB (United Kingdom)

2010-06-15

Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R and D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic-ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic-metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg{sub HM} discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW{sub (th)} loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life

Spent fuel reprocessing options

International Nuclear Information System (INIS)

2008-08-01

The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

Conceptual design of the fusion-driven subcritical system FDS-I

International Nuclear Information System (INIS)

Wu, Y.; Zheng, S.; Zhu, X.; Wang, W.; Wang, H.; Liu, S.; Bai, Y.; Chen, H.; Hu, L.; Chen, M.; Huang, Q.; Huang, D.; Zhang, S.; Li, J.; Chu, D.; Jiang, J.; Song, Y.

2006-01-01

The fusion-driven subcritical system (named FDS-I) was previously proposed as an intermediate step toward the final application of fusion energy. A conceptual design of the FDS-I is presented, which consists of the fusion neutron driver with relatively easy-achieved plasma parameters, and the He-gas/liquid lithium-lead Dual-cooled subcritical Waste Transmutation (DWT) blanket used to transmute long-lived radioactive wastes and to generate energy on the basis of self-sustainable fission and fusion fuel cycle. An overview of the FDS-I is given and the specifications of the design analysis are summarized

Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

International Nuclear Information System (INIS)

Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

1986-01-01

Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

Fuel utilization improvement in PWRs using the denatured 233U-Th cycle

International Nuclear Information System (INIS)

Jones, H.M.; Schwenk, G.A.; Toops, E.C.; Yotinen, V.O.

1980-06-01

A number of changes in PWR core design and/or operating strategy were evaluated to assess the fuel utilization improvement achievable by their implementation in a PWR using thorium-based fuel and operating in a recycle mode. The reference PWR for this study was identical to the B and W Standard Plant except that the fuel pellets were of denatured ( 233 U/ 238 U-Th)O 2 . An initial scoping study identified the three most promising improvement concepts as (1) a very tight lattice, (2) thorium blankets, and (3) ThO 2 rods placed in available guide tubes. A conceptual core design incorporating these changes was then developed, and the fuel utilization of this modified design was compared with that of the reference case

U.S. -- EC fuel cycle study: Background document to the approach and issues

Energy Technology Data Exchange (ETDEWEB)

Cantor, Robin; Russell, Lee; Krupnick, Alan; Smith, Hilary; Schaffhauser, Jr., A.; Barnthouse, Larry; Cada, Glen; Kroodsma, Roger; Turner, Robb; Easterly, Clay; Jones, Troyce; Burtraw, Dallas; Harrington, Winston; Freeman, A. Myrick

1992-11-01

In February 1991, DOE and the Commission of the European Communities (EC), signed a joint statement regarding the external costs of fuel cycles. This 18-month agreement committed their respective organizations to develop a comparative analytical methodology and to develop the best range of estimates of external costs from secondary sources'' for eight fuel cycles and four conservation options. In our study, a fuel cycle is defined as the series of physical and chemical processes and activities that are required to generate electricity from a specific fuel or resource. This foundation phase of the study is primarily limited to developing and demonstrating methods for estimating impacts and their monetized value, what we term damages'' or benefits,'' leaving aside the extent to which such damages have been internalized. However, Appendix C provides the conceptual framework for evaluating the extent of internalization. This report is a background document to introduce the study approach and to discuss the major conceptual and practical issues entailed by the incremental damage problem. As a background document, the report seeks to communicate an overview of the study and the important methodological choices that were made to conduct the research. In successive sections of the report, the methodological tools used in the study are discussed; the ecological and health impacts are reviewed using the coal fuel cycle as a reference case; and, in the final chapter, the methods for valuing impacts are detailed.

Conceptual Model of Iodine Behavior in the Subsurface at the Hanford Site

Energy Technology Data Exchange (ETDEWEB)

Truex, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lee, Brady D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Johnson, Christian D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qafoku, Nikolla [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Szecsody, James E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kyle, Jennifer E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tfaily, Malak M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle MV [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Saunders, Danielle L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lawter, Amanda R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Oostrom, Martijn L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tartakovsky, Guzel D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Leavy, Ian I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); McElroy, Erin M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Appriou, Delphine [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sahajpal, Rahul [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Carroll, Matthew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chu, Rosalie K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cordova, Elsa [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Last, George V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lee, Hope [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kaplan, Daniel I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Garcia, Whitney L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kerisit, Sebastien N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qafoku, Odeta [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bowden, Mark E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Toyoda, Jason G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Plymale, Andrew E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-09-01

Isotopes of iodine were generated during plutonium production within the nine production reactors at the U.S. Department of Energy Hanford Site. The short half-life 131I that was released from the fuel into the atmosphere during the dissolution process (when the fuel was dissolved) in the Hanford Site 200 Area is no longer present at concentrations of concern in the environment. The long half-life 129I generated at the Hanford Site during reactor operations was (1) stored in single-shell and double-shell tanks, (2) discharged to liquid disposal sites (e.g., cribs and trenches), (3) released to the atmosphere during fuel reprocessing operations, or (4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129I to the subsurface have resulted in several large, though dilute, plumes in the groundwater. There is also 129I remaining in the vadose zone beneath disposal or leak locations. The fate and transport of 129I in the environment and potential remediation technologies are currently being studied as part of environmental remediation activities at the Hanford Site. A conceptual model describing the nature and extent of subsurface contamination, factors that control plume behavior, and factors relevant to potential remediation processes is needed to support environmental remedy decisions. Because 129I is an uncommon contaminant, relevant remediation experience and scientific literature are limited. In addition, its behavior in subsurface is different from that of other more common and important contaminants (e.g., U, Cr and Tc) in terms of sorption (adsorption and precipitation), and aqueous phase species transformation via redox reactions. Thus, the conceptual model also needs to both describe known contaminant and biogeochemical process information and identify aspects about which additional information is needed to effectively support remedy decisions.

Technical meeting on 'Review of solid and mobile fuels for partitioning and transmutation systems'. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

The topics covered during the Meeting were divided into two Sessions. Session 1 - Qualification of Solid and Mobile Fuels delt with: Neutronic, fuel and material properties of a molten salt transmuter; and Preliminary analysis of transmutation fuels for KALIMER. Session 2 - Reactor Physics and Safety Characteristics of Transmutation Systems based on Solid and Mobile Fuel Types included the following: Activity in NEA for P and T area; IAEA activities in the area of partitioning and transmutation; The R and D activity in Brazil: A conceptual fast energy amplifier ADS cooled by helium double stata Th/U fuel cycle; Closed fuel cycle and contemporary tendencies of the nuclear facilities development; Current Russian activities in P and T area; Pyrochemical reprocessing and nuclear spent fuel disposal project; Fuel selection criteria specific for double stratum minor actinide burners.

Optimum design of a self-supported power transmission tower type 2M2 for 230 kW; Diseno optimo de una torre de transmision autosoportada tipo 2M2, para 230 kW

Energy Technology Data Exchange (ETDEWEB)

Espejel Valdez, Eduardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Millan Monje, Alejandro; Honk Hernandez, Wenceslao [Comision Federal de Electricidad, (CFE), Mexico, D. F. (Mexico)

1986-12-31

A group of specialists from the Comision Federal de Electricidad (CFE) and of the Instituto de Investigaciones Electricas (IIE) analyzes the design of various types of power transmission towers in their application stage, in order to enhance them. In this article the structural characteristics of the self-supported tower type 2M2 for 230 kV of two circuits and of suspension, are presented. [Espanol] Un grupo de especialistas de la Comision Federal de Electricidad (CFE) y del Instituto de Investigaciones Electricas (IIE) analiza el diseno de diversos tipos de torres de transmision en su etapa de aplicacion, con el fin de mejorarlos. En este articulo se presentan las caracteristicas estructurales de la torre autosoportada tipo 2M2 para 230 kV, de dos circuitos y de suspension.

Optimum design of a self-supported power transmission tower type 2M2 for 230 kW; Diseno optimo de una torre de transmision autosoportada tipo 2M2, para 230 kW

Energy Technology Data Exchange (ETDEWEB)

Espejel Valdez, Eduardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Millan Monje, Alejandro; Honk Hernandez, Wenceslao [Comision Federal de Electricidad, (CFE), Mexico, D. F. (Mexico)

1985-12-31

A group of specialists from the Comision Federal de Electricidad (CFE) and of the Instituto de Investigaciones Electricas (IIE) analyzes the design of various types of power transmission towers in their application stage, in order to enhance them. In this article the structural characteristics of the self-supported tower type 2M2 for 230 kV of two circuits and of suspension, are presented. [Espanol] Un grupo de especialistas de la Comision Federal de Electricidad (CFE) y del Instituto de Investigaciones Electricas (IIE) analiza el diseno de diversos tipos de torres de transmision en su etapa de aplicacion, con el fin de mejorarlos. En este articulo se presentan las caracteristicas estructurales de la torre autosoportada tipo 2M2 para 230 kV, de dos circuitos y de suspension.

Heuristic rules analysis on the fuel cells design using greedy search;Analisis de reglas heuristicas en el diseno de celdas de combustible usando busqueda greedy

Energy Technology Data Exchange (ETDEWEB)

Ortiz, J. J.; Castillo, J. A.; Montes, J. L.; Hernandez, J. L., E-mail: juanjose.ortiz@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2009-10-15

This work approaches the study of one of the heuristic rules of fuel cells design for boiling water nuclear reactors. This rule requires that the minor uranium enrichment is placed in the corners of the fuel cell. Also the search greedy is applied for the fuel cells design where explicitly does not take in count this rule, allowing the possibility to place any uranium enrichment with the condition that it does not contain gadolinium. Results are shown in the quality of the obtained cell by search greedy when it considers the rule and when not. The cell quality is measured with the value of the power pick factor obtained, as well as of the neutrons multiplication factor in an infinite medium. Cells were analyzed with 1 and 2 gadolinium concentrations low operation conditions at 120% of the nominal power of the reactors of the nuclear power plant of Laguna Verde. The results show that not to consider the rule in cells with a single gadolinium concentration, it has as repercussion that the greedy search has a minor performance. On the other hand with cells of two gadolinium concentrations, the performance of the greedy search was better. (Author)

The external costs of the nuclear fuel cycle: implementation in France

International Nuclear Information System (INIS)

Dreicer, M.; Tort, V.; Margerie, H.

1995-08-01

In 1991 the European Community and the US Department of Energy initiated a joint research project to assess the external costs of fuel cycles used to generate electricity. The intention of this project, called the EC-US External Costs of Fuel Cycles Project (ECFC), was to develop a conceptual approach, consistent methodology and identify future research in the assessment of the externalities. A second phase of the project continued in Europe (with a new name ''ExternE'') and expanded to include the implementation of the consistent methodology in various EC countries. This report presents the final results of the French Implementation for the nuclear fuel cycle. (author). 37 refs., 11 figs., 24 tabs

Comparative of fuel cycle cost for light water nuclear power plants

International Nuclear Information System (INIS)

Kocic, A.; Dimitrijevic, Z.

1978-01-01

Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

The external costs of the nuclear fuel cycle: implementation in France

Energy Technology Data Exchange (ETDEWEB)

Dreicer, M.; Tort, V.; Margerie, H.

1995-08-01

In 1991 the European Community and the US Department of Energy initiated a joint research project to assess the external costs of fuel cycles used to generate electricity. The intention of this project, called the EC-US External Costs of Fuel Cycles Project (ECFC), was to develop a conceptual approach, consistent methodology and identify future research in the assessment of the externalities. A second phase of the project continued in Europe (with a new name ``ExternE``) and expanded to include the implementation of the consistent methodology in various EC countries. This report presents the final results of the French Implementation for the nuclear fuel cycle. (author). 37 refs., 11 figs., 24 tabs.

Preliminary conceptual design and cost estimation for Korea Advanced Pyroprocessing Facility Plus (KAPF+)

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il, E-mail: nwiko@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Ho Hee, E-mail: nhhlee@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Choi, Sungyeol, E-mail: csy@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Sung-Ki, E-mail: sgkim1@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Byung Heung, E-mail: b.h.park@ut.ac.kr [Department of Chemical and Biological Engineering, Korea National University of Transportation, 50 Daehak-ro, Chungju-si, Chungbuk, 380-702 (Korea, Republic of); Lee, Hyo Jik, E-mail: hyojik@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, In Tae, E-mail: nitkim@kaeri.re.kr [Department of Chemical and Biological Engineering, Korea National University of Transportation, 50 Daehak-ro, Chungju-si, Chungbuk, 380-702 (Korea, Republic of); Lee, Han Soo, E-mail: hslee5@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2014-10-01

Highlights: • Conceptual design is created for a pilot pyroprocessing plant treating PWR spent fuel. • Pilot-scale design is based on a capacity of 400 tHM/yr with 60 years lifetime. • All individual processes are integrated into a single system from feed to products. • Overall facility design is developed for a pilot pyroprocessing plant. • Unit process cost is estimated for pyroprocessing with uncertainties. - Abstract: Korea has developed pyroprocessing technology as a potential option for recycling spent fuels (SFs) from pressurized water reactors (PWRs). The pyroprocessing consists of various key unit processes and a number of research activities have been focused on each process. However, to realize the whole pyroprocessing concept, there is a critical need for integrating the individual developments and addressing a material flow from feed to final products. In addition, the advancement on overall facility design is an indispensable aspect for demonstration and commercialization of the pyroprocessing. In this study, a facility named as Korea Advanced Pyroprocess Facility Plus (KAPF+) is conceptualized with a capacity of 400 tHM/yr. The process steps are categorized based on their own characteristics while the capacities of process equipment are determined based on the current technical levels. The facility concept with a site layout of 104,000 m{sup 2} is developed by analyzing the operation conditions and materials treated in each process. As an economic approach to the proposed facility, the unit cost (781 $/kgHM denominated in 2009 USD) for KAPF+ is also analyzed with the conceptual design with preliminary sensitivity assessments including decontamination and decommissioning costs, a discount rate, staffing costs, and plant lifetime. While classifying and describing cost details of KAPF+, this study compares the unit cost of KAPF+ treating PWR SF to that of the pyroprocessing facility treating sodium-cooled fast reactor (SFR) SF.

The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

International Nuclear Information System (INIS)

Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2003-03-01

In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc

The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2003-03-01

In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.

Conceptual structures in practice

CERN Document Server

Hitzler, Pascal

2009-01-01

Exploring fundamental research questions, Conceptual Structures in Practice takes you through the basic yet nontrivial task of establishing conceptual relations as the foundation for research in knowledge representation and knowledge mining. It includes contributions from leading researchers in both the conceptual graph and formal concept analysis (FCA) communities.This accessible, self-contained book begins by providing the formal background in FCA and conceptual graphs. It then describes various software tools for analysis and computation, including the ToscanaJ suite. Written by the origina

Numerical simulations of a full-scale polymer electrolyte fuel cell with analysing systematic performance in an automotive application

International Nuclear Information System (INIS)

Park, Heesung

2015-01-01

Highlights: • A 3-D full-scale fuel cell performance is numerically simulated. • Generated and consumed power in the system is affected by operating condition. • Systematic analysis predicts the net power of conceptual PEFC stack. - Abstract: In fuel cell powered electric vehicles, the net power efficiency is a critical factor in terms of fuel economy and commercialization. Although the fuel cell stack produces enough power to drive the vehicles, the transferred power to the power train could be significantly reduced due to the power consumption to operate the system components of air blower and cooling module. Thus the systematic analysis on the operating condition of the fuel cell stack is essential to predict the net power generation. In this paper numerical simulation is conducted to characterize the fuel cell performance under various operating conditions. Three dimensional and full-scale fuel cell of the active area of 355 cm 2 is numerically modelled with 47.3 million grids to capture the complexities of the fluid dynamics, heat transfer and electrochemical reactions. The proposed numerical model requires large computational time and cost, however, it can be powerful to reasonably predict the fuel cell system performance at the early stage of conceptual design without requiring prototypes. Based on the model, it has been shown that the net power is reduced down to 90% of the gross power due to the power consumption of air blower and cooling module

Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

Energy Technology Data Exchange (ETDEWEB)

Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1997-11-01

A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

Fuel cell propulsion for urban duty vehicles: Bavarian fuel cell bus project

International Nuclear Information System (INIS)

Wurster, R.; Altmann, M.; Sillat, D.; Kalk, K. W.; Hammerschmidt, A.; Stuehler, W.; Holl, E.

1998-01-01

Following a feasibility study and a detailed specification phase, the realization of a fuel cell city bus prototype was started in autumn 1996. The project is a joint development effort of Siemens, MAN and Linde, which receives a 50 % funding by the Bavarian State Ministry for Economic Affairs, Transport and Technology (BStMWVT) in the context of the Hydrogen Initiative Bavaria. An MAN low-floor bus will be equipped with the components for a fuel cell drive system. The PEM fuel cell is developed by the power generation division of Siemens. Four fuel cell modules deliver a total electrical output of 120 kW to the two electric motors, which are linked by a summation gearbox by the Siemens Transportation Systems Division. MAN Technologie AG is responsible for the compressed hydrogen storage system allowing for a driving range of more than 250 km, while Linde AG takes care of the hydrogen periphery and delivers the hydrogen for the test operation scheduled for the beginning of the year 2000. Project coordination is done by Ludwig-Boelkow System-technik GmbH. The project is divided into four phases. The conceptual design phase is scheduled to last until the end of 1997. The partly overlapping system integration phase will end in the first quarter of 1999. The subsequent test and commissioning phase will prepare the test operation at the beginning of 2000 with a bus operator yet to be defined. (author)

Advanced nuclear fuel cycles - Main challenges and strategic choices

International Nuclear Information System (INIS)

Le Biez, V.; Machiels, A.; Sowder, A.

2013-01-01

A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness

Advanced nuclear fuel cycles - Main challenges and strategic choices

Energy Technology Data Exchange (ETDEWEB)

Le Biez, V. [Corps des Mines, 35 bis rue Saint-Sabin, F-75011 Paris (France); Machiels, A.; Sowder, A. [Electric Power Research Institute, Inc. 3420, Hillview Avenue, Palo Alto, CA 94304 (United States)

2013-07-01

A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

Design study on advanced nuclear fuel recycle system. Conceptual design study of recycle system using molten salt

International Nuclear Information System (INIS)

Kasai, Y.; Kakehi, I.; Moro, T.; Higashi, T.; Tobe, K.; Kawamura, F.; Yonezawa, S.; Yoshiuji, T.

1998-10-01

Advanced recycle system engineering group of OEC (Oarai Engineering Center) has being carried out a design study of the advanced nuclear fuel recycle system using molten salt (electro-metallurgical process). This system is aiming for improvements of fuel cycle economy and reduction of environmental burden (MA recycles, Minimum of radioactive waste disposal), and also improvement of safety and nuclear non-proliferation. This report describes results of the design study that has been continued since December 1996. (1) A design concept of the advanced nuclear fuel recycle system, that is a module type recycles system of pyrochemical reprocessing and fuel re-fabrication was studied. The module system has advantage in balance of Pu recycle where modules are constructed in coincidence with the construction plan of nuclear power plants, and also has flexibility for technology progress. A demonstration system, minimum size of the above module, was studies. This system has capacity of 10 tHM/y and is able to demonstrate recycle technology of MOX fuel, metal fuel and nitride fuel. (2) Each process of the system, which are pyrochemical electrorefining system, cathode processor, de-cladding system, waste disposal system, etc., were studied. In this study, capacity of an electrorefiner was discussed, and vitrification experiment of molten salt using lead-boric acid glass was conducted. (3) A hot cell system and material handling system of the demonstration system was studied. A robot driven by linear motor was studied for the handling system, and an arrangement plan of the cell system was made. Criticality analysis in the cell system and investigation of material accountancy system of the recycle plant were also made. This design study will be continued in coincidence with design study of reactor and fuel, aiming to establish the concept of FBR recycle system. (author)

Investigation of small scale sphere-pac fuel fabrication plant with external gelation process

International Nuclear Information System (INIS)

Maekawa, Kazuhiko; Yoshimura, Tadahiro; Kikuchi, Toshiaki; Hoshino, Yasushi; Munekata, Hideki; Shimizu, Makoto

2005-02-01

In feasibility studies on commercialized FBR cycle system, comprehensive system investigation and properties evaluation for candidate FBR cycle systems have been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation of small scale sphere-pac fuel fabrication plant with external gelation process was conducted. Until last fiscal year, equipment layout in cells and overall layout design of the 200t-HM/y scale fuel fabrication plant were conducted as well as schematical design studies on main equipments in gelation and reagent recovery processes of the plant. System property data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In this fiscal year, the processes from vibropacking to fuel assemblies storage were added to the investigation range, and a conceptual design of whole fuel fabrication plant was studied as well as deepening the design study on main equipments. The conceptual design study was mainly conducted for small 50t-HM/y scale plant and a revising investigation was done for 200t-HM/y scale plant. Taking the planed comparative evaluation with pellet fuel fabrication system into account, design of equipments which should be equivalent with pellet system, especially in post-vibropacking processes, were standardized in each system. Based on these design studies, system properties data concerning economics and environmental burden reduction of the plant was also acquired. In comparison with existing design, the cell height was lowered on condition that plug type pneumatic system was adopted and fuel fabrication building was downsized by applying rationalized layout design of pellet system to post-vibropacking processes. Reduction of reagent usage at gelation process and rationalization of sintering and O/M controlling processes etc., are foremost tasks. (author)

Fabrication of metallic channel-containing UO2 fuels

International Nuclear Information System (INIS)

Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

2004-01-01

The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

A study relating to the conceptual design and cost estimates of the Saskatchewan Power Corporation lignite beneficiation process

Energy Technology Data Exchange (ETDEWEB)

Granatstein, D L; Champagne, P J; Szladow, A J

1982-06-01

The purpose is to provide technology for upgrading lignite to a cleaner fuel that is capable of improving utility plant operation and is more acceptable to the non-utility sector. Beneficiated lignite can provide a middle alternative between low-grade and high-grade fuels. The results for phase IV include a conceptual design for a lignite beneficiation plant and a marketing study for the product. The beneficiated lignite is cheaper to transport, easier to handle, cleaner, allows better plant operation with higher availability, reduced slagging and fouling, reduced sulphur, and lower operating and maintenance costs. 8 refs., 6 figs., 7 tabs.

A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks

International Nuclear Information System (INIS)

Lima, Alan M.M. de; Schirru, Roberto

2005-01-01

A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

Conceptual designs of power tokamak-type thermonuclear reactors

International Nuclear Information System (INIS)

Shejndlin, A.E.; Nedospasov, A.V.

1978-01-01

Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

Simulated coal-gas fueled carbonate fuel cell power plant system verification. Final report, September 1990--June 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

This report summarizes work performed under U.S. Department of Energy, Morgantown Energy Technology Center (DOE/METC) Contract DE-AC-90MC27168 for September 1990 through March 1995. Energy Research Corporation (ERC), with support from DOE, EPRI, and utilities, has been developing a carbonate fuel cell technology. ERC`s design is a unique direct fuel cell (DFC) which does not need an external fuel reformer. An alliance was formed with a representative group of utilities and, with their input, a commercial entry product was chosen. The first 2 MW demonstration unit was planned and construction begun at Santa Clara, CA. A conceptual design of a 10OMW-Class dual fuel power plant was developed; economics of natural gas versus coal gas use were analyzed. A facility was set up to manufacture 2 MW/yr of carbonate fuel cell stacks. A 100kW-Class subscale power plant was built and several stacks were tested. This power plant has achieved an efficiency of {approximately}50% (LHV) from pipeline natural gas to direct current electricity conversion. Over 6,000 hours of operation including 5,000 cumulative hours of stack operation were demonstrated. One stack was operated on natural gas at 130 kW, which is the highest carbonate fuel cell power produced to date, at 74% fuel utilization, with excellent performance distribution across the stack. In parallel, carbonate fuel cell performance has been improved, component materials have been proven stable with lifetimes projected to 40,000 hours. Matrix strength, electrolyte distribution, and cell decay rate have been improved. Major progress has been achieved in lowering stack cost.

Disposal criticality analysis for aluminum-based DOE fuels

International Nuclear Information System (INIS)

Davis, J.W.; Gottlieb, P.

1997-11-01

This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels

A Conceptual Mirror

DEFF Research Database (Denmark)

Badie, Farshad

2017-01-01

The multilevel interactions between a mentor and her/his learner could exchange various conceptions between them that are supported by their own conceptualisations. Producing the own realisation of a world and developing it in the context of interactions could be said to be the most valuable prod...... will analyse the logical dependencies between learner and men- tor and will check their reflectional symmetrical relationship in a conceptual mirror. The conceptual mirror is a phenomenon that represents the meeting point of the mentor’s and the learner’s conceptual knowledge....

The advanced neutron source three-element-core fuel grading

International Nuclear Information System (INIS)

Gehin, J.C.

1995-01-01

The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered

Development and application of a conceptual approach for defining high-level waste

International Nuclear Information System (INIS)

Croff, A.G.; Forsberg, C.W.; Kocher, D.C.; Cohen, J.J.; Smith, C.F.; Miller, D.E.

1986-01-01

This paper presents a conceptual approach to defining high-level radioactive waste (HLW) and a preliminary quantitative definition obtained from an example implementation of the conceptual approach. On the basis of the description of HLW in the Nuclear Waste Policy Act of 1982, we have developed a conceptual model in which HLW has two attributes: HLW is (1) highly radioactive and (2) requires permanent isolation via deep geologic disposal. This conceptual model results in a two-dimensional waste categorization system in which one axis, related to ''requires permanent isolation,'' is associated with long-term risks from waste disposal and the other axis, related to ''highly radioactive,'' is associated with short-term risks from waste management and operations; this system also leads to the specification of categories of wastes that are not HLW. Implementation of the conceptual model for defining HLW was based primarily on health and safety considerations. Wastes requiring permanent isolation via deep geologic disposal were defined by estimating the maximum concentrations of radionuclides that would be acceptable for disposal using the next-best technology, i.e., greater confinement disposal (GCD) via intermediate-depth burial or engineered surface structures. Wastes that are highly radioactive were defined by adopting heat generation rate as the appropriate measure and examining levels of decay heat that necessitate special methods to control risks from operations in a variety of nuclear fuel-cycle situations. We determined that wastes having a power density >200 W/m 3 should be considered highly radioactive. Thus, in the example implementation, the combination of maximum concentrations of long-lived radionuclides that are acceptable for GCD and a power density of 200 W/m 3 provides boundaries for defining wastes that are HLW

Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-12-15

Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process.

Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

International Nuclear Information System (INIS)

Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won

2008-12-01

Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process

An Opportunity for Hydrogen Fueled Supersonic Airliners

Directory of Open Access Journals (Sweden)

Alex Forbes

2011-02-01

Full Text Available This paper takes a new look at the prospects for developing supersonic civil airliners, considering global demographics, climate change issues, fuel prices and technological advances. Dramatic changes have occurred in the demographics, economics, and market intensity of the Eastern Hemisphere since the 1990s. Carbon reduction imperatives provide a major incentive to invest in developing hydrogen-fueled airliners. The “point-to-point” air route architecture has proved viable with long range mid-size airliners. With a cruise Mach number of 1.4, a large number of destinations become viable for overland supersonic flight. A conceptual design process is used to estimate cost per seat mile for a range of hydrocarbon and hydrogen fuel costs. An argument based on the ideal shape for minimal wave drag, estimates the drag penalty from using hydrogen. Viable aircraft geometries are shown to exist, that match the theoretical ideal shape, showing that the drag estimate is achievable. Conservative design arguments and market estimates suggest that hydrogen-fueled airliners can achieve seat-mile costs low enough to open a large worldwide market and justify a viable fleet size.

Process Design and Economics for the Conversion of Algal Biomass to Biofuels: Algal Biomass Fractionation to Lipid-and Carbohydrate-Derived Fuel Products

Energy Technology Data Exchange (ETDEWEB)

None

2014-09-11

The U.S. Department of Energy (DOE) promotes the production of a range of liquid fuels and fuel blendstocks from biomass feedstocks by funding fundamental and applied research that advances the state of technology in biomass production, conversion, and sustainability. As part of its involvement in this program, the National Renewable Energy Laboratory (NREL) investigates the conceptual production economics of these fuels. This includes fuel pathways from lignocellulosic (terrestrial) biomass, as well as from algal (aquatic) biomass systems.

Development of a spent fuel management technology research and test facility

Energy Technology Data Exchange (ETDEWEB)

Park, S W; Noh, S K; Lee, J S. and others

1997-12-01

This study was intended to develop concept for a pilot-scale remote operation facility for longer term management of spent fuel and therefrom to provide technical requirement for later basic design of the facility. Main scope of work for the study was to revise the past (1990) conceptual design in functions, scale, hot cell layout, etc. based on user requirements. Technical reference was made to the PKA facility in Germany, through collaboration with appropriate partner, to elaborate the design and requirements. A simulator of the conceptual design was also developed by use of virtual reality technique by 3-D computer graphics for equipment and building. (author). 18 tabs., 39 figs

Hydroxide Self-Feeding High-Temperature Alkaline Direct Formate Fuel Cells.

Science.gov (United States)

Li, Yinshi; Sun, Xianda; Feng, Ying

2017-05-22

Conventionally, both the thermal degradation of the anion-exchange membrane and the requirement of additional hydroxide for fuel oxidation reaction hinder the development of the high-temperature alkaline direct liquid fuel cells. The present work addresses these two issues by reporting a polybenzimidazole-membrane-based direct formate fuel cell (DFFC). Theoretically, the cell voltage of the high-temperature alkaline DFFC can be as high as 1.45 V at 90 °C. It has been demonstrated that a proof-of-concept alkaline DFFC without adding additional hydroxide yields a peak power density of 20.9 mW cm -2 , an order of magnitude higher than both alkaline direct ethanol fuel cells and alkaline direct methanol fuel cells, mainly because the hydrolysis of formate provides enough OH - ions for formate oxidation reaction. It was also found that this hydroxide self-feeding high-temperature alkaline DFFC shows a stable 100 min constant-current discharge at 90 °C, proving the conceptual feasibility. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Conceptual overview and preliminary risk assessment of cryogen use in deep underground mine production

Science.gov (United States)

Sivret, J.; Millar, D. L.; Lyle, G.

2017-12-01

This research conducts a formal risk assessment for cryogenic fueled equipment in underground environments. These include fans, load haul dump units, and trucks. The motivating advantage is zero-emissions production in the subsurface and simultaneous provision of cooling for ultra deep mine workings. The driving force of the engine is the expansion of the reboiled cryogen following flash evaporation using ambient temperature heat. The cold exhaust mixes with warm mine air and cools the latter further. The use of cryogens as ‘fuel’ leads to much increased fuel transport volumes and motivates special considerations for distribution infrastructure and process including: cryogenic storage, distribution, handling, and transfer systems. Detailed specification of parts and equipment, numerical modelling and preparation of design drawings are used to articulate the concept. The conceptual design process reveals new hazards and risks that the mining industry has not yet encountered, which may yet stymie execution. The major unwanted events include the potential for asphyxiation due to oxygen deficient atmospheres, or physical damage to workers due to exposure to sub-cooled liquids and cryogenic gases. The Global Minerals Industry Risk Management (GMIRM) framework incorporates WRAC and Bow-Tie techniques and is used to identify, assess and mitigate risks. These processes operate upon the competing conceptual designs to identify and eliminate high risk options and improve the safety of the lower risk designs.

Fuel density effect on parameter of reactivity coefficient of the Innovative Research Reactor core

International Nuclear Information System (INIS)

Rokhmadi; Tukiran S

2013-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR) and then as an exchanger for old research reactor. The aim of this research is to complete RRI core design data as a requirement for design license. Calculation done is to get the RRI core reactivity coefficients with 5 x 5 core configuration and 20 MW of power, has more than 40 days cycle of length. The RRI core reactivity coefficient calculation is done for new U-"9Mo-Al fuel with variation of densities. The calculation is done by using WIMSD-5B and BATAN-FUEL computer codes. The result of calculation for conceptual design showed that the equilibrium RRI core with 5 x 5 configuration, 450 g, 550 g and 700 g of fuel loadings have negative reactivity coefficients of fuel temperature, moderator temperature, void fraction and density of moderator but the values of the reactivities are very variation. This results has met the safety criteria for RRI core conceptual design. (author)

Social and environmental impact of the bioclimatic design; Impacto social y ambiental del diseno bioclimatico

Energy Technology Data Exchange (ETDEWEB)

Lopez P, J. Manuel A; Morrillon G, David; Rodriguez V, Luis [Instituto de Ingenieria de la UNAM, Mexico, D.F. (Mexico)

2000-07-01

This article is about two important scopes of the Bioclimatical Design (BD), the Social scope and the environmental scope. As an introduction, it is justified the importance of this topic through the association of the BD in the context of sustainability with our country's economic model. The analysis is done under the perspective of effects of the BD on social and environmental scopes. In social scope is used the Cost and Benefit Analysis according to social evaluation. In environmental scope, effects in water, air and ground are studied, but the measurement is done according to the emissions reduction of CO{sub 2} on atmosphere because of the earn on energy consumption. The results of the measurement of this effects are related with economical costs and benefits, which may show a general view about economical, social and environmental scopes. As conclusions, interest results are show and this is proposed to incorporate them in the economical politics. [Spanish] En este articulo se tocan dos aspectos importantes del Diseno Bioclimatico (DB), el Impacto Social y el Impacto Ambiental. En la introduccion se justifica la importancia del tema, asociando el DB en el contexto de la sustentabilidad, con el modelo economico que se ha imperado en nuestro pais. El estudio se realiza bajo la perspectiva de los impactos que el DB tiene en el aspecto social y en el ambiental. En lo social, se recurre a la metodologia del Analisis Costo-Beneficio para evaluacion Social, en lo ambiental, aunque se tocan los puntos de impacto en aire, agua y uso de suelo, la medicion del impacto se realiza con base en la disminucion de emisiones de CO{sub 2} a la atmosfera por el ahorro en el consumo de energia electrica. Los resultados de la medicion de impactos se cruzan con los costos y beneficios economicos, para presentar un panorama completo en cuanto a aspectos economicos, sociales y ambientales. En las conclusiones se rescatan los resultados de interes y se proponen las inserciones de

A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

International Nuclear Information System (INIS)

Lima, Alan M.M. de; Schirru, Roberto; Carvalho da Silva, Fernando; Medeiros, Jose Antonio Carlos Canedo

2008-01-01

The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem

A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

Energy Technology Data Exchange (ETDEWEB)

Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

2008-09-15

The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

Radionuclides and isotopes release of spent fuel matrix. Conceptual and mathematical models of wastes behaviour; Liberacion de los radionucleidos e isotopos estables contenidos en la matriz del combustible. Modelo conceptual y modelo matematico del comportamiento del residuo

Energy Technology Data Exchange (ETDEWEB)

Cera, E.; Merino, J.; Bruno, J.

2000-07-01

We have developed a conceptual and numerical model to calculate release of selected radionuclides from spent fuel under repository condition. This has been done in the framework of the Enresa 2000 performance assessment exercise. The model has been developed based on kinetic mass balance equations in order to study the evolution of the spent fuel water interface as a function of time. Several processes have been kinetically modelled: congruent dissolution, radioactive decay, ingrowth and water turnover in the gap. The precipitation/redissolution of secondary solid phases has been taken into account from a thermodynamic point of view. Both approaches have been coupled and the resulting equations solved for a number of radionuclides in both, a conservative and realistic approach. The results show three distinct groups of radionuclides based on their release behaviour: a first group is composed of radioisotopes of highly insoluble elements (e. g., Pu, Am, Pd) whose concentration in the gap is mainly controlled by their solubility and therefore their evolution is identical in both cases. Secondly, a set of radionuclides from soluble elements under these conditions (e. g., I, Cs, Ra) show concentrations kinetically controlled, decreasing with time following the congruent dissolution trend. Their release concentrations are one order of magnitude larger in the conservative case than in the realistic case. Finally, a third group has been identified (e. g., Se, Th, Cm) where a mixed behaviour takes place: initially their solubility limiting phases control their concentration in the gap but the situation reverts to a kinetic control as the chemical conditions change and the secondary precipitates become totally dissolved. The fluxes of the different radionuclides are also given as an assessment of the source term in the performance assessment. (Author)

Thorium utilization program progress report for January 1, 1974--June 30, 1975. [Reprocessing; refabrication; recycle fuel irradiations

Energy Technology Data Exchange (ETDEWEB)

Lotts, A.L.; Kasten, P.R.

1976-05-01

Work was carried out on the following: HTGR reprocessing development and pilot plant, refabrication development and pilot plant, recycle fuel irradiations, engineering and economic studies, and conceptual design of a commercial recycle plant. (DLC)

SOLID STATE ENERGY CONVERSION ALLIANCE (SECA) SOLID OXIDE FUEL CELL PROGRAM

Energy Technology Data Exchange (ETDEWEB)

Nguyen Minh; Jim Powers

2003-10-01

This report summarizes the work performed for April 2003--September 2003 reporting period under Cooperative Agreement DE-FC26-01NT41245 for the U.S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Solid State Energy Conversion Alliance (SECA) Solid oxide Fuel Cell Program''. During this reporting period, the conceptual system design activity was completed. The system design, including strategies for startup, normal operation and shutdown, was defined. Sealant and stack materials for the solid oxide fuel cell (SOFC) stack were identified which are capable of meeting the thermal cycling and degradation requirements. A cell module was tested which achieved a stable performance of 0.238 W/cm{sup 2} at 95% fuel utilization. The external fuel processor design was completed and fabrication begun. Several other advances were made on various aspects of the SOFC system, which are detailed in this report.

Preliminary conceptual studies of REX 2000

International Nuclear Information System (INIS)

Merchie, F.; Baas, C.; Ballagny, A.; Chagrot, M.; Farny, G.; Barnier, M.; Pattou, A.

1993-01-01

Nuclear R and D programs are, to some extent, completely dependent on research reactors availability. In France and others european countries, the major materials testings reactors were built in the sixties and are consequently ageing and reaching the end of their life, some of them having already been shut down. A situation with not a single large research reactor available in first half of next century cannot be imagined, given all the benefits drawn from the use of research reactors. The CEA has therefore started to evaluate the needs for neutron sources in the next four or five decades so as to design the most suitable new facilities to take over from the existing ones. REX 2000 is a new dedicated reactor project intended to meet the needs for fuels and materials testings after the year 2000. The preliminary conceptual studies which have been carried out along the last 18 months are presented and commented. (author)

Plasma Heating and Fueling Department

International Nuclear Information System (INIS)

Alsmiller, R.G.; Baer, M.B.; Barber, G.C.

1976-01-01

ORMAK neutral injection systems were studied on a test stand, and the results of these studies, coupled with the upgrading of power supplies and electrical systems and with the addition of high-power modulators, allowed increased neutral beam power for ORMAK. A conceptual design of a TFTR neutral beam system was completed, and an initial scoping of an EPR neutral injection system was performed. A test facility for the development of 60-kV, 60-A ion sources was completed and is operational. A 15-cm-diameter ion source was developed and was operated up to 30 A at 27 kV. A 20-cm version of this source, having design parameters of 60 A, 40 A, and 300 msec, was built for PLT. The continuing source evolution was aided by the development of a plasma source discharge model. A pellet fueling program was initiated with the goal of developing pellet fueling devices capable of meeting future power reactor requirements

Dry refabrication technology development of spent nuclear fuel

International Nuclear Information System (INIS)

Park, Geun Il; Lee, J. W.; Song, K. C.

2012-04-01

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed

Dry refabrication technology development of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Park, Geun Il; Lee, J. W.; Song, K. C.; and others

2012-04-15

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed.

Development of a strategy for the management of PBMR spent fuel in South Africa

Energy Technology Data Exchange (ETDEWEB)

Smith, S.W., E-mail: Schalk.Smith@necsa.co.z [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa); Bredell, P.J. [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa)

2010-10-15

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would form an integral part of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development.

Conceptual design for irradiation device used to irradiate experimental LFR fuel element in TRIGA reactor, ACPR zone

International Nuclear Information System (INIS)

Ioan, M.

2013-01-01

The paper presents the main steps followed to conceive a small, versatile and rather cheep irradiation device used for irradiation of an experimental fuel element, specific for Lead cooled Fast Reactor (LFR), adapted to TRIGA reactor, ACPR zone. This device must be instrumented with at least 4 thermocouples and a pressure transducer. The fuel element (150 mm fuel pellets column) will be immersed in maximum 0.350 kg pure hot lead (400 deg C). The system has three protection barriers, as follows: first is the fuel tube, second is the lead container (maximum 20 mm inner diameter) and third is the external container (maximum 180 mm outside diameter). Before the reactor pulse, the temperature of the lead is set at the prescribed value using an electrical heater (300 W), coil on the second barrier. Outside the second barrier a very good thermal insulation is provided. (authors)

Conceptual Design of a 150-Passenger Civil Tiltrotor

Science.gov (United States)

Costa, Guillermo

2012-01-01

The conceptual design of a short-haul civil tiltrotor aircraft is presented. The concept vehicle is designed for runway-independent operations to increase the capacity of the National Airspace System without the need for increased infrastructure. This necessitates a vehicle that is capable of integrating with conventional air traffic without interfering with established flightpaths. The NASA Design and Analysis of Rotorcraft software was used to size the concept vehicle based on the mission requirements of this market. The final configuration was selected based upon performance metrics such as acquisition and maintenance costs, fuel fraction, empty weight, and required engine power. The concept presented herein has a proposed initial operating capability date of 2035, and is intended to integrate with conventional air traffic as well as proposed future air transportation concepts.

Spent fuel management in Japan - Facts and prospects

International Nuclear Information System (INIS)

Nagano, K.

2002-01-01

This paper discusses recent developments and future issues related to spent fuel management in Japan. With increasing pressure of spent fuel discharge from the power plants in operation and, in contrast, uncertainties in their processing and management services, spent fuel storage in short and medium terms has been receiving the highest priority in nuclear policy discussions in Japan. While small-scale interim storage devices, as well as capacity expansion (re-racking, etc.) and shared uses of existing devices, are introduced at number of power stations, large scale AFR (away from reactor) 'Storage of Recycle Fuel Resources' is expected to come in a medium and long-run. Commercial operation of 'Storage of Recycle Fuel Resources' is allowed its way, as the bill of amendment to the law for regulation of nuclear power reactors and other nuclear-related activities has passed in the Diet. In the meantime, the Atomic Energy Commission has launched working group discussions for revision of 'The Long-term Program of Research, Development and Utilization of Nuclear Energy' to be completed in 2000. This revision is hoped to set up a stage of national debate of nuclear policy, which might lead to fill conceptual gaps between bodies promoting nuclear development and general public. The author's attempt to illustrate the role of storage in spent fuel management is also presented from a theoretical point of view. (author)

Thorium fuels for heavy water reactors. Romanian experience

International Nuclear Information System (INIS)

Glodeanu, F.; Mirion, I.; Mehedinteanu, S.; Balan, V.

1984-01-01

The renewed interest in thorium fuel cycle due to the increased demand for fissile materials has resulted in speeding up the related research and development activities. For heavy water reactors the thorium cycles, especially SSET, are very promising and many efforts are made to demonstrate their feasibility. In our country, at INPR, the research and development activity has been initiated in the following areas: the conceptual design of thorium bearing fuel elements; fuel modelling; nuclear grade thorium dioxide powder technology; mixed oxide fuel technology. In the design area, the key factors in performance limitation, especially at extended burnup have been accounted and different remedies proposed. An irradiation programme has been settled and will start this year. The modelling activities are focused on mixed oxide behaviour and material data measurements are in progress. In the nuclear grade thorium powder technology area, a good piece of work has been done to develop an integrated technology for monasite processing (thorium being a by-product in lanthanides extraction). As regards the mixed oxide fuel technology, efforts have been made to obtain (ThU)O 2 pellets with good homogeneity and high density at different compositions. Besides the mixing powders route, other non-conventional technologies for refabrication like: microspheres, pellet impregnation and clay extrusion are studied. Experimental fuel rods for irradiation testing have been manufactured. (author)

Mixed Reload Design Using MOX and UOX Fuel Assemblies

International Nuclear Information System (INIS)

Ramon, Ramirez Sanchez J.; Perry, R.T.

2002-01-01

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

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

Conceptual engineering design studies of 1985-era commercial VTOL and STOL transports that utilize rotors

Science.gov (United States)

Magee, J. P.; Clark, R. D.; Widdison, C. A.

1975-01-01

Conceptual design studies are summarized of tandem-rotor helicopter and tilt-rotor aircraft for a short haul transport mission in the 1985 time frame. Vertical takeoff designs of both configurations are discussed, and the impact of external noise criteria on the vehicle designs, performance, and costs are shown. A STOL design for the tilt-rotor configuration is reported, and the effect of removing the vertical takeoff design constraints on the design parameters, fuel economy, and operating cost is discussed.

Characteristics of fast reactor core designs and closed fuel cycle

International Nuclear Information System (INIS)

Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

2007-01-01

On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

SAF-BRET-FMEF: a developmental LMR fuel cycle facility

International Nuclear Information System (INIS)

Stradley, J.G.; Yook, H.R.; Gerber, E.W.; Lerch, R.E.; Rice, L.H.

1985-01-01

The SAF-BRET-FMEF complex represents a versatile fuel cycle facility for processing LMR fuel. While originally conceived for processing FFTF and CRBRP fuel, it represents a facility where LMR fuel from the first generation of innovative LMRs could be processed. The cost of transporting fuel from the LMR to the Hanford site would have to be assessed when the LMR site is identified. The throughput of BRET was set at 15 MTHM/yr during conceptual design of the facility, a rate which was adequate to process all of the fuel from FFTF and fuel and blanket material from CRBRP. The design is currently being reevaluated to see if BRET could be expanded to approx.35 MTHM/yr to process fuel and blanket material from approx.1300 MWe generating capacity of the innovative LMRs. This expanded throughput is possible by designing the equipment for an instantaneous throughput of 0.2 MTHM/d, and by selected additional modifications to the facility (e.g., expansion of shipping and receiving area, and addition of a second entry tunnel transporter), and by the fact that the LMR fuel assemblies contain more fuel than the FFTF assemblies (therefore, fewer assemblies must be handled for the same throughput). The estimated cost of such an expansion is also being assessed. As stated previously, the throughput of SAF and Fuel Assembly could be made to support typical LMRs at little additional cost. The throughput could be increased to support the fuel fabrication requirements for 1300 MWe generating capacity of the innovative LMRs. This added capacity may be achieved by increasing the number of operating shifts, and is affected by variables such as fuel design, fuel enrichment, and plutonium isotopic composition

Study for the selection of a supplementary spent fuel storage facility for KANUPP

International Nuclear Information System (INIS)

Ahmed, W.; Iqbal, M.J.; Arshad, M.

1999-01-01

Steps taken for construction of the spent fuel facility of Karachi Nuclear Power Plant (KANUPP) are the following: choice of conceptual design and site selection; preliminary design and preparation of Preliminary Safety Analysis Report (PSAR); Construction of the facility and preparation of PSAR; testing/commissioning and loading of the storage facility. Characterisation of the spent fuel is essential for design of the storage facility. After comparison of various storage types, it seems that construction of dry storage facility based on concrete canisters at KANUPP site is a suitable option to enhance the storage capacity

Chinese FBR program and its first FR conceptual design

International Nuclear Information System (INIS)

Xu Mi

1991-01-01

As the preliminary study results, The long term strategy of FBR development in China would be divided into three steps: Experimental Fast Reactor, Modulized Fast Breeder Reactor and Large Fast Breeder Reactor. The design requirements, main technical selections and design boundary conditions for the first step, experimental fast reactor, which is named FFR (First Fast Reactor) in China have been given in the paper. The conceptual design of FFR, based on core neutronics, core subassemblies thermo-hydraulics, reactor block and mechanics, fuel handling system, main heat transfer system and steam-electricity generation system has been completed and also presented in this paper. Finally, the R and D program of FFR which is carring on and will be done is briefly introduced. (author)

Bootstrapping: Una teoría explicativa del cambio conceptual Bootstrapping: A theory for conceptual change

Directory of Open Access Journals (Sweden)

José Antonio Castorina

2005-12-01

Full Text Available El presente artículo expone la teoría explicativa propuesta por Carey para el cambio conceptual. Primeramente, se plantea la cuestión de la reorganización conceptual en la psicología cognitiva y la posición de Carey. En segundo lugar, se ponen de relieve las condiciones epistémica que deben cumplir las "teorías" infantiles para que la reestructuración conceptual sea posible, así como los modos que adopta esta última. En tercer lugar, se muestran los resultados de investigaciones que verifican el cambio conceptual entre teorías infantiles de biología intuitiva. En cuarto lugar, se plantean las dificultades de otras teorías del cambio conceptual, para luego formular los rasgos del mecanismo alternativo de bootstrapping y su pertinencia para interpretrar los datos de las indagaciones mencionadas. Finalmente, se evalúan la originalidad de la teoría del bootstrpping en el escenario de los debates contemporáneos. Muy especialmente, se esboza una posible aproximación con las tesis dialécticas de Piaget.This paper examines the Carey's theory of conceptual change. First, it describes the conceptual reorganization in cognitive psychology and the author position. Second, the epistemic conditions that children "theories" should fulfil to make conceptual restructuring possible, as well as the ways adopted by the latter, are analyzed. In third place, findings of researches testing the conceptual change among biology intuitive children theories are explained. Subsequently, it discusses the difficulties other theories of conceptual change present, in order to state features of bootstrapping as an alternative mechanism and its relevance for the interpretation of abovementioned researches results. Finally, it evaluates the originality of "bootstrapping" theory in the scene of contemporary debates. It particularly outlines a possible approach to Piaget's dialectic theses.

Numerical study of radial stepwise fuel load reshuffling traveling wave reactor

International Nuclear Information System (INIS)

Zhang Dalin; Zheng Meiyin; Tian Wenxi; Qiu Suizheng; Su Guanghui

2015-01-01

Traveling wave reactor is a new conceptual fast breeder reactor, which can adopt natural uranium, depleted uranium and thorium directly to realize the self sustainable breeding and burning to achieve very high fuel utilization fraction. Based on the mechanism of traveling wave reactor, a concept of radial stepwise fuel load reshuffling traveling wave reactor was proposed for realistic application. It was combined with the typical design of sodium-cooled fast reactors, with which the asymptotic characteristics of the inwards stepwise fuel load reshuffling were studied numerically in two-dimension. The calculated results show that the asymptotic k_e_f_f parabolically varies with the reshuffling cycle length, while the burnup increases linearly. The highest burnup satisfying the reactor critical condition is 38%. The power peak shifts from the fuel discharging zone (core centre) to the fuel uploading zone (core periphery) and correspondingly the power peaking factor decreases along with the reshuffling cycle length. In addition, at the high burnup case the axial power distribution close to the core centre displays the M-shaped deformation. (authors)

Conceptual Combination During Sentence Comprehension

Science.gov (United States)

Swinney, David; Love, Tracy; Walenski, Matthew; Smith, Edward E.

2008-01-01

This experiment examined the time course of integration of modifier-noun (conceptual) combinations during auditory sentence comprehension using cross-modal lexical priming. The study revealed that during ongoing comprehension, there is initial activation of features of the noun prior to activation of (emergent) features of the entire conceptual combination. These results support compositionality in conceptual combination; that is, they indicate that features of the individual words constituting a conceptual combination are activated prior to combination of the words into a new concept. PMID:17576278

SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

International Nuclear Information System (INIS)

D. A. Thomas

1996-01-01

The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report

Technology development goals for automotive fuel cell power systems. Final report

Energy Technology Data Exchange (ETDEWEB)

James, B.D.; Baum, G.N.; Kuhn, I.F. Jr. [Directed Technologies, Inc., Arlington, VA (United States)

1994-08-01

This report determines cost and performance requirements for Proton Exchange Membrane (PEM) fuel cell vehicles carrying pure H{sub 2} fuel, to achieve parity with internal combustion engine (ICE) vehicles. A conceptual design of a near term FCEV (fuel cell electric vehicle) is presented. Complete power system weight and cost breakdowns are presented for baseline design. Near term FCEV power system weight is 6% higher than ICE system, mid-term FCEV projected weights are 29% lower than ICE`s. There are no inherently high-cost components in FCE, and at automotive production volumes, near term FCEV cost viability is closer at hand than at first thought. PEM current vs voltage performance is presented for leading PEM manufacturers and researchers. 5 current and proposed onboard hydrogen storage techniques are critically compared: pressurized gas, cryogenic liquid, combined pressurized/cryogenic, rechargeable hydride, adsorption. Battery, capacitor, and motor/controller performance is summarized. Fuel cell power system component weight and cost densities (threshold and goal) are tabulated.

Integrated international safeguards concepts for fuel reprocessing

International Nuclear Information System (INIS)

Hakkila, E.A.; Gutmacher, R.G.; Markin, J.T.; Shipley, J.P.; Whitty, W.J.; Camp, A.L.; Cameron, C.P.; Bleck, M.E.; Ellwein, L.B.

1981-12-01

This report is the fourth in a series of efforts by the Los Alamos National Laboratory and Sandia National Laboratories, Albuquerque, to identify problems and propose solutions for international safeguarding of light-water reactor spent-fuel reprocessing plants. Problem areas for international safeguards were identified in a previous Problem Statement (LA-7551-MS/SAND79-0108). Accounting concepts that could be verified internationally were presented in a subsequent study (LA-8042). Concepts for containment/surveillance were presented, conceptual designs were developed, and the effectiveness of these designs was evaluated in a companion study (SAND80-0160). The report discusses the coordination of nuclear materials accounting and containment/surveillance concepts in an effort to define an effective integrated safeguards system. The Allied-General Nuclear Services fuels reprocessing plant at Barnwell, South Carolina, was used as the reference facility

A conceptual design of catalytic gasification fuel cell hybrid power plant with oxygen transfer membrane

Science.gov (United States)

Shi, Wangying; Han, Minfang

2017-09-01

A hybrid power generation system integrating catalytic gasification, solid oxide fuel cell (SOFC), oxygen transfer membrane (OTM) and gas turbine (GT) is established and system energy analysis is performed. In this work, the catalytic gasifier uses steam, recycled anode off-gas and pure oxygen from OTM system to gasify coal, and heated by hot cathode off-gas at the same time. A zero-dimension SOFC model is applied and verified by fitting experimental data. Thermodynamic analysis is performed to investigate the integrated system performance, and system sensitivities on anode off-gas back flow ratio, SOFC fuel utilization, temperature and pressure are discussed. Main conclusions are as follows: (1) System overall electricity efficiency reaches 60.7%(HHV) while the gasifier operates at 700 °C and SOFC at 850 °C with system pressure at 3.04 bar; (2) oxygen enriched combustion simplify the carbon-dioxide capture process, which derives CO2 of 99.2% purity, but results in a penalty of 6.7% on system electricity efficiency; (3) with SOFC fuel utilization or temperature increasing, the power output of SOFC increases while GT power output decreases, and increasing system pressure can improve both the performance of SOFC and GT.

CONCEPTUAL METAPHORS IN BASKETBALL DISCOURSE

Directory of Open Access Journals (Sweden)

Reda Toleikienė

2013-04-01

Full Text Available This article presents part of the research performed within the scope of the national project “Conceptual Metaphors in Public Discourse,”1 financed by the Research Council of Lithuania. The aim of the present paper is to analyze conceptual metaphors in the discourse of the European Basketball Championship which took place in Kaunas, Lithuania in 2011, as well as to determine the source concepts. The analysis allows certain features of the images which are used while conceptualizing the entities related to basketball to be described. The metaphorical collocations drawn from the Lithuanian language corpora and web portals (www.delfi.lt and www.lrytas.lt were selected and analyzed from 31 August 2011 to 18 September 2011. A conceptual metaphor is defined as an interaction of two conceptual fields (source and target concepts. On the basis of the analyzed conceptual sayings, the reconstructed conceptual metaphors proved that the most prolific metaphors are of war, ontology, and scale. In basketball discourse, the war metaphor is characterized by the fact that the image of sport is war is supplemented by other source concepts (e.g.,a person, a building, a thing, a material, a scale. The features of two or sometimes even three source concepts are ascribed to the target concept.

Conceptual study of dry storage method for spent fuel assemblies based on honeycomb concrete overpack (COP). Phase 1

International Nuclear Information System (INIS)

Hida, Yoshio; Hayashi, Shigeki; Katsuyama, Yoshiaki; Hashimoto, Hirohide; Murata, Takashi

2017-01-01

The amount of spent fuel assemblies currently stored in Japan is approximately 15,000 tU. Most of these are stored in storage pools, although dry storage method will be safer, as was revealed in the accident of the Fukushima Daiichi Nuclear Power Plant. In addition, Japan has established a national policy of the nuclear fuel cycle. All spent fuel assemblies are designated for reprocessing. However, the reprocessing plant in Japan is currently under regulatory review for compliance with newly established safety standards. Beyond this, shortfalls in its processing capacity mean interim storage facilities for spent fuel are required. The Tokyo Electric Power Company Holdings, Incorporated and the Japan Atomic Power Company are currently building an interim dry storage facility with a storage capacity of 5,000 tU in Aomori Prefecture, while Chubu Electric Power Company, Inc. is currently building a dry storage facility with a storage capacity of 400 tU in the Hamaoka Nuclear Power Station. These facilities consist of earthquake-resistant buildings and dry storage casks. Within the buildings, metal transportable storage casks loaded with spent fuel assemblies are placed vertically with spaces between the casks and supported by earthquake-proof measures that prevent toppling or other movement. These structures entail significant cost and construction efforts. At the Fukushima Daiichi Nuclear Power Plant, a temporary dry storage facility has been built within the premises to store spent fuel generated during decommissioning. Part of this facility is already in operation. Here, each metal cask containing spent fuel is mounted on an earthquake-resistant concrete mat, which is anchored to the ground. Each cask is enclosed in a concrete box for additional radiation shielding, and the casks are spaced at intervals. This approach requires a large plot of land. The dry storage method for spent fuel presented here does not require a building. The dry metal casks containing spent

Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

1975-06-01

The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

Design Research as Conceptual Designing

DEFF Research Database (Denmark)

Ylirisku, Salu; Jacucci, Giulio; Sellen, Abigail

2015-01-01

define conceptual designing as a constructive framing and re-framing activity, which is mediated by and targeted at the creation of new design concepts. Conceptual designing as an approach is valuable for addressing the fuzziness and ambiguity typical of research that explores novel areas with new...... partners, methods and resources. It is by no means a new phenomenon, and the main contribution of the article is the clarification of conceptual designing as a particular approach to designing and researching. The approach embraces openness, resource-construction and collaboration. We conclude...... that conceptual designing can be especially useful in research and design projects that bring different kinds of people, organizations, technologies and domains together into the forming of new well-founded proposals for development. The presentation of conceptual designing in this paper is written...

German Approach to Spent Fuel Management

International Nuclear Information System (INIS)

Jussofie, A.; Graf, R.; Filbert, W.

2010-01-01

in the pilot conditioning facility. The conditioning mode was taken over for the subsequent BSK3 concept which varies in the canister for the fuel rods and the emplacement technology. To prove the principle feasibility of this concept aboveground demonstration tests have been performed by 2009. Investigations on a further concept intended for the direct disposal of spent fuel assemblies without conditioning have just begun. Conceptual differences may have an impact on the safeguards approach. With the end of the moratorium the political situation will change and further steps related to the final disposal of heat-generating waste could be decided. (author)

Statistical mechanical analysis of LMFBR fuel cladding tubes

International Nuclear Information System (INIS)

Poncelet, J.-P.; Pay, A.

1977-01-01

The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. (Auth.)

Exact and conceptual repetition dissociate conceptual memory tests: problems for transfer appropriate processing theory.

Science.gov (United States)

McDermott, K B; Roediger, H L

1996-03-01

Three experiments examined whether a conceptual implicit memory test (specifically, category instance generation) would exhibit repetition effects similar to those found in free recall. The transfer appropriate processing account of dissociations among memory tests led us to predict that the tests would show parallel effects; this prediction was based upon the theory's assumption that conceptual tests will behave similarly as a function of various independent variables. In Experiment 1, conceptual repetition (i.e., following a target word [e.g., puzzles] with an associate [e.g., jigsaw]) did not enhance priming on the instance generation test relative to the condition of simply presenting the target word once, although this manipulation did affect free recall. In Experiment 2, conceptual repetition was achieved by following a picture with its corresponding word (or vice versa). In this case, there was an effect of conceptual repetition on free recall but no reliable effect on category instance generation or category cued recall. In addition, we obtained a picture superiority effect in free recall but not in category instance generation. In the third experiment, when the same study sequence was used as in Experiment 1, but with instructions that encouraged relational processing, priming on the category instance generation task was enhanced by conceptual repetition. Results demonstrate that conceptual memory tests can be dissociated and present problems for Roediger's (1990) transfer appropriate processing account of dissociations between explicit and implicit tests.

Remotex and servomanipulator needs in nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

Garin, J.

1981-01-01

Work on the conceptual design of a pilot-scale plant for reprocessing breeder reactor fuels is being performed at Oak Ridge National Laboratory. The plant design will meet all current federal regulations for repocessing plants and will serve as prototype for future production plants. A unique future of the concept is the incorporation of totally remote operation and maintenance of the process equipment within a large barn-like hot cell. This approach, caled Remotex, utilizes servomanipulators coupled with television viewing to extend man's capabilities into the hostile cell environment. The Remotex concept provides significant improvements for fuel reprocessing plants and other nuclear facilities in the areas of safeguarding nuclear materials, reducing radiation exposure, improving plant availability, recovering from unplanned events, and plant decommissioning

Implementation of the Bee Colony Optimization method for the design of fuel cells; Implementacion del metodo Bee Colony Optimization para el diseno de celdas de combustible

Energy Technology Data Exchange (ETDEWEB)

Esquivel E, J.; Ortiz S, J. J., E-mail: jaime.esquivel@fi.uaemex.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2011-11-15

The present work shows the results obtained after applying the Bee Colony Optimization algorithm in the design of fuel cells for a BWR. The algorithm that is implemented, works following the behavior that have the bees when pollinating a flowers field. The bees carry out an exhaustive analysis in the cell, so they leave generating diverse configurations where different fuel bars are placed with different uranium enrichments to reach a value mean, with a specific number of gadolinium bars. The behavior of the generated cell is evaluated by means of the use of the commercial code CASMO-4, which shows the variables that allow fixing if the cell fulfills the requirements. Such variables are the local potential peak factor and the neutrons multiplication factor in an infinite medium. (Author)

DDE-MURR Status Report of Conceptual Design Activities

Energy Technology Data Exchange (ETDEWEB)

N.E. Woolstenhulme; R.B. Nielson; M.H. Sprenger; G.K. Housley

2012-09-01

The Design Demonstration Experiment for the University of Missouri Research Reactor (DDE-MURR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the MURR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in a 200mm channel at the Belgium Reactor 2. At the time this report was prepared the resources for furthering DDE design work were expected to be postponed. As such, the conceptual design effort to date is summarized herein in order to provide the status of key objectives, notable results, and provisions for future design work. These demonstrate that the DDE-MURR design effort is well on the path to producing a suitable irradiation experiment, but also exhibits several challenges for which timely resolution is recommend in order to facilitate success of the irradiation campaign and ultimate conversion of the MURR.

DDE-MITR Status Report of Conceptual Design Activities

Energy Technology Data Exchange (ETDEWEB)

N.E. Woolstenhulme; R.B. Nielson; J.D. Wiest; J.W. Nielsen; G.A. Roth; S.D. Snow

2012-09-01

The Design Demonstration Experiment for the Massachusetts Institute of Technology Reactor (DDE-MITR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the MITR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in the Advanced Test Reactor center flux trap. At the time this report was prepared the resources for furthering DDE design work were expected to be postponed. As such, the conceptual design effort to date is summarized herein in order to provide the status of key objectives, notable results, and provisions for future design work. These demonstrate that the DDE-MITR design effort is well on the path to producing a suitable irradiation experiment, but also exhibits several challenges for which timely resolution is recommend in order to facilitate success of the irradiation campaign and ultimate conversion of the MITR.

DDE-NBSR Status Report of Conceptual Design Activities

Energy Technology Data Exchange (ETDEWEB)

N.E. Woolstenhulme; R.B. Nielson; B.P. Durtschi; C.R. Glass; G.A. Roth; D.T. Clark

2012-09-01

The Design Demonstration Experiment for the National Bureau of Standard Reactor (DDE-NBSR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the NBSR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in the Advanced Test Reactor center flux trap. At the time this report was prepared the resources for furthering DDE design work were expected to be postponed. As such, the conceptual design effort to date is summarized herein in order to provide the status of key objectives, notable results, and provisions for future design work. These demonstrate that the DDE-NBSR design effort is well on the path to producing a suitable irradiation experiment, but also exhibits several challenges for which timely resolution is recommend in order to facilitate success of the irradiation campaign and ultimate conversion of the NBSR.

Computational program to design heat pumps by compression (ciclo 1.0); Programa computacional para diseno de bombas de calor por compresion (ciclo 1.0)

Energy Technology Data Exchange (ETDEWEB)

De Alba Rosano, Mauricio [CIE, UNAM, Temixco, Morelos (Mexico)

2000-07-01

A new computational program has been developed in order to design single stage compression heat pumps. This software, named CICLO 1.0 allows the design of water-water, water-air, air-water and air-air heat pumps, for industrial and residential applications. CICLO 1.0 simulates three types of compressors: reciprocating, screw and scroll. Also has a data base created with REFPROP software which includes eleven refrigerants. The condenser and evaporator simulation includes global conductance (UA) determination, and when one or both are shell and tube's type, this software shows the even number of tube passes by shell. The software determines the best compressor and refrigerant setup taking the COP as a parameter; in order to obtain this, is necessary to know the inlet/outlet conditions of the fluid to be heated, the inlet conditions of the fluid that gives heat, and the electric motor efficiency that drives the compressor. The afforded results by CICLO 1.0 are: operation conditions from compression cycle, that means, pressures and temperatures at the inlet/outlet from every heat pump component are determined: as well as refrigerant mass flux, COP, power required by compressor, volumetric and isentropic efficiencies, heat exchangers global conductance and more data. CICLO 1.0 has been executed with heat pump data that nowadays are operating, and the results from the simulation have been very similar each other with data reported from operational facilities. [Spanish] Se ha desarrollado un nuevo programa computacional para el diseno de bombas de calor por compresion de vapor de una sola etapa. Este programa, CICLO 1.0, permite el diseno de bombas de calor de tipo: agua-agua, agua-aire, aire-agua y aire-aire, que se utilicen para aplicaciones industriales, de servicios y residenciales. CICLO 1.0 simula tres tipos de compresores: reciprocante, de tornillo y scroll: cuenta con una base de datos de refrigerantes creada con el programa REFPROP la cual incluye once

Material control and accountability aspects of safeguards for the USA 233U/Th fuel recycle plant

International Nuclear Information System (INIS)

Carpenter, J.A. Jr.; McNeany, S.R.; Angelini, P.; Holder, N.D.; Abraham, L.

1978-01-01

The materials control and accountability aspects of the reprocessing and refabrication of a conceptual large-scale HTGR fuel recycle plant have been discussed. Two fuel cycles were considered. The traditional highly enriched uranium cycle uses an initial or makeup fuel element with a fissile enrichment of 93% 235 U. The more recent medium enriched uranium cycle uses initial or makeup fuel elements with a fissile enrichment less than 20% 235 U. In both cases, 233 U bred from the fertile thorium is recycled. Materials control and accountability in the plant will be by means of a real-time accountability method. Accountability data will be derived from monitoring of total material mass through the processes and a system of numerous assays, both destructive and nondestructive

Monitored Retrievable Storage conceptual system study: metal storage casks

International Nuclear Information System (INIS)

Unterzuber, R.; Cross, T.E.; Krasicki, B.R.

1983-08-01

A description of the metal cask storage facility concept is presented with the operations required to handle the spent fuel or high-level wastes and transuranic wastes. A generic Receiving and Handling Facility, provided by PNL, has been used for this study. Modifications to the storage delivery side of the handling facility, necessary to couple the Receiving and Handling Facility with the storage facility, are described. The equipment and support facilities needed for the storage facility are also described. Two separate storage facilities are presented herein: one for all spent fuel storage, and one for storage of high-level waste (HLW) and transuranic waste (TRU). Each facility is described for the capacities and rates defined by PNL in the Concept Technical Performance Criteria and Base Assumptions (see Table 1.3-1). Estimates of costs and time-distributions of expenditures have been developed to construct, operate, and decommission the conceptual MRS facilities in mid-1983 dollars, for the base cases given using the cost categories and percentages provided by PNL. Cost estimates and time-distributions of expenditures have also been developed to expand the facility throughput rate from 1800 MTU to 3000 MTU, and to expand the facility storage capacity from 15,000 MTU to 72,00 MTU. The life cycle cost of the facility for the bounding cases of all spent fuel and all HLW and TRU, using the time-distributions of costs developed above and assuming a two percent per year discount rate, are also presented. 3 references, 16 figures, 18 tables

Revealing conceptual understanding of international business

NARCIS (Netherlands)

Sue Ashley; Dr. Harmen Schaap; Prof.Dr. Elly de Bruijn

2017-01-01

This study aims to identify an adequate approach for revealing conceptual understanding in higher professional education. Revealing students’ conceptual understanding is an important step towards developing effective curricula, assessment and aligned teaching strategies to enhance conceptual

Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

International Nuclear Information System (INIS)

Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

2002-12-01

In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

Development of intelligent code system to support conceptual design of nuclear reactor core

International Nuclear Information System (INIS)

Kugo, Teruhiko; Nakagawa, Masayuki; Tsuchihashi, Keichiro

1997-01-01

An intelligent reactor design system IRDS has been developed to support conceptual design of new type reactor cores in the fields of neutronics, thermal-hydraulics and fuel behavior. The features of IRDS are summarized as follows: 1) a variety of computer codes to cover various design tasks relevant to 'static' and 'burnup' problems are implemented, 2) all the information necessary to the codes implemented is unified in a data base, 3) several data and knowledge bases are referred to in order to proceed design process efficiently for non-expert users, 4) advanced man-machine interface to communicate with the system through an interactive and graphical user interface is equipped and 5) a function to search automatically a design window, which is defined as a feasible parameter range to satisfy design requirement and criteria is employed to support the optimization or satisfication process. Applicability and productivity of the system are demonstrated by the design study of fuel pin for new type FBR cores. (author)

Neutronics and Thermal Hydraulics Analysis of a Conceptual Ultra-High Temperature MHD Cermet Fuel Core for Nuclear Electric Propulsion

Directory of Open Access Journals (Sweden)

Jian Song

2018-04-01

Full Text Available Nuclear electric propulsion (NEP offers unique advantages for the interplanetary exploration. The extremely high conversion efficiency of magnetohydrodynamics (MHD conversion nuclear reactor makes it a highly potential space power source in the future, especially for NEP systems. Research on ultra-high temperature reactor suitable for MHD power conversion is performed in this paper. Cermet is chosen as the reactor fuel after a detailed comparison with the (U,ZrC graphite-based fuel and mixed carbide fuel. A reactor design is carried out as well as the analysis of the reactor physics and thermal-hydraulics. The specific design involves fuel element, reactor core, and radiation shield. Two coolant channel configurations of fuel elements are considered and both of them can meet the demands. The 91 channel configuration is chosen due to its greater heat transfer performance. Besides, preliminary calculation of nuclear criticality safety during launch crash accident is also presented. The calculation results show that the current design can meet the safety requirements well.

DDE-MURR Status Report of Conceptual Design Activities

Energy Technology Data Exchange (ETDEWEB)

N.E. Woolstenhulme; R.B. Nielson; M.H. Sprenger; G.K. Housley

2013-09-01

The Design Demonstration Experiment for the University of Missouri Research Reactor (DDE-MURR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the MURR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in a 200mm channel at the Belgium Reactor 2 (BR2). Revision 0 of this report was prepared at the end of government fiscal year 2012 when most of the resources for furthering DDE design work were expected to be postponed. Hence, the conceptual design efforts were summarized to provide the status of key objectives, notable results, and provisions for future design work. Revision 1 of this report was prepared at the end of fiscal year 2013 in order to include results from a neutronic study performed by BR2, to incorporate further details that had been achieved in the engineering sketches of the irradiation devices, and to provide an update of the DDE-MURR campaign in relation to program objectives and opportunities for its eventual irradiation. These updates were purposed to bring the DDE-MURR conceptual design to level of maturity similar to that of the other two DDE efforts (DDE-MITR and DDE-NBSR). This report demonstrates that the DDE-MURR design effort is well on the path to producing a suitable irradiation experiment, but also puts forth several recommendations in order to facilitate success of the irradiation campaign.

Conceptual and non-conceptual repetition priming in category exemplar generation: Evidence from bilinguals.

Science.gov (United States)

Francis, Wendy S; Fernandez, Norma P; Bjork, Robert A

2010-10-01

One measure of conceptual implicit memory is repetition priming in the generation of exemplars from a semantic category, but does such priming transfer across languages? That is, do the overlapping conceptual representations for translation equivalents provide a sufficient basis for such priming? In Experiment 1 (N=96) participants carried out a deep encoding task, and priming between languages was statistically reliable, but attenuated, relative to within-language priming. Experiment 2 (N=96) replicated the findings of Experiment 1 and assessed the contributions of conceptual and non-conceptual processes using a levels-of-processing manipulation. Words that underwent shallow encoding exhibited within-language, but not between-language, priming. Priming in shallow conditions cannot therefore be explained by incidental activation of the concept. Instead, part of the within-language priming effect, even under deep-encoding conditions, is due to increased availability of language-specific lemmas or phonological word forms.

Revealing Conceptual Understanding of International Business

Science.gov (United States)

Ashley, Sue; Schaap, Harmen; de Bruijn, Elly

2017-01-01

This study aims to identify an adequate approach for revealing conceptual understanding in higher professional education. Revealing students' conceptual understanding is an important step towards developing effective curricula, assessment and aligned teaching strategies to enhance conceptual understanding in higher education. Essays and concept…

Development of core technology for research reactors using plate type fuels

International Nuclear Information System (INIS)

Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

2009-12-01

Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

Energy Technology Data Exchange (ETDEWEB)

Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

Conceptual Design and Cost Estimate of a Subsonic NASA Testbed Vehicle (NTV) for Aeronautics Research

Science.gov (United States)

Nickol, Craig L.; Frederic, Peter

2013-01-01

A conceptual design and cost estimate for a subsonic flight research vehicle designed to support NASA's Environmentally Responsible Aviation (ERA) project goals is presented. To investigate the technical and economic feasibility of modifying an existing aircraft, a highly modified Boeing 717 was developed for maturation of technologies supporting the three ERA project goals of reduced fuel burn, noise, and emissions. This modified 717 utilizes midfuselage mounted modern high bypass ratio engines in conjunction with engine exhaust shielding structures to provide a low noise testbed. The testbed also integrates a natural laminar flow wing section and active flow control for the vertical tail. An eight year program plan was created to incrementally modify and test the vehicle, enabling the suite of technology benefits to be isolated and quantified. Based on the conceptual design and programmatic plan for this testbed vehicle, a full cost estimate of $526M was developed, representing then-year dollars at a 50% confidence level.

Climate change adaptation, damages and fossil fuel dependence. An RETD position paper on the costs of inaction

Energy Technology Data Exchange (ETDEWEB)

Katofsky, Ryan; Stanberry, Matt; Hagenstad, Marca; Frantzis, Lisa

2011-07-15

The Renewable Energy Technology Deployment (RETD) agreement initiated this project to advance the understanding of the ''Costs of Inaction'', i.e. the costs of climate change adaptation, damages and fossil fuel dependence. A quantitative estimate was developed as well as a better understanding of the knowledge gaps and research needs. The project also included some conceptual work on how to better integrate the analyses of mitigation, adaptation, damages and fossil fuel dependence in energy scenario modelling.

Overlooking the Conceptual Framework

Science.gov (United States)

Leshem, Shosh; Trafford, Vernon

2007-01-01

The conceptual framework is alluded to in most serious texts on research, described in some and fully explained in few. However, examiners of doctoral theses devote considerable attention to exploring its function within social science doctoral vivas. A literature survey explores how the conceptual framework is itself conceptualised and explained.…

Analysis of fuel management in the KIPT neutron source facility

Energy Technology Data Exchange (ETDEWEB)

Zhong Zhaopeng, E-mail: zzhong@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2011-05-15

Research highlights: > Fuel management of KIPT ADS was analyzed. > Core arrangement was shuffled in stage wise. > New fuel assemblies was added into core periodically. > Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is {approx}360 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

Development of a strategy for the management of PBMR spent fuel in South Africa - HTR2008-58047

International Nuclear Information System (INIS)

Smith, S. W.; Bredell, P. J.; Meyer, W. C. M. H.

2008-01-01

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development. (authors)

Design of a nuclear fuel rod support grid using axiomatic design

International Nuclear Information System (INIS)

Song, Kee Nam; Yoon, Kyung Ho; Kang, Byung Soo; Park, Gyung Jin; Choi, Sung Kyoo

2002-01-01

Recently, much attention is imposed on the design of the fuel assemblies in the Pressurized Light Water Reactor (PWR). Spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water, and maintains a coolable geometry from the external impact loads. In this research, a new shape of the spacer grid is designed by the axiomatic approach. The Independence axiom is utilized for the design. For conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy FRs in sequence. Overall configuration and shapes are determined in this process. Detail design is carried out based on the result of the axiomatic design. For the detail design, the system performances are evaluated by using linear and nonlinear finite element analysis. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design

How can conceptual schemes change teaching?

Science.gov (United States)

Wickman, Per-Olof

2012-03-01

Lundqvist, Almqvist and Östman describe a teacher's manner of teaching and the possible consequences it may have for students' meaning making. In doing this the article examines a teacher's classroom practice by systematizing the teacher's transactions with the students in terms of certain conceptual schemes, namely the epistemological moves, educational philosophies and the selective traditions of this practice. In connection to their study one may ask how conceptual schemes could change teaching. This article examines how the relationship of the conceptual schemes produced by educational researchers to educational praxis has developed from the middle of the last century to today. The relationship is described as having been transformed in three steps: (1) teacher deficit and social engineering, where conceptual schemes are little acknowledged, (2) reflecting practitioners, where conceptual schemes are mangled through teacher practice to aid the choices of already knowledgeable teachers, and (3) the mangling of the conceptual schemes by researchers through practice with the purpose of revising theory.

Pre-conceptual design of a spent PWR fuel disposal container

International Nuclear Information System (INIS)

Choi, Jong Won; Cho, Dong Keun; Lee, Yang; Choi, Heui Joo; Lee, Jong Youl

2005-01-01

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid and bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert. the Maximum Von Mises stress from the 102 cm container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by ∼20 tons

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Preliminary conceptual design of the blanket and power conversion system for the Mirror Hybrid Reactor

International Nuclear Information System (INIS)

Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

1978-01-01

A conceptual design of a commercial Mirror Hybrid Reactor, optimized for 239 Pu production, has been completed. This design is the product of a joint effort by Lawrence Livermore Laboratory and General Atomic Company, and follows directly from earlier work on the Mirror Hybrid. This paper describes the blanket and power conversion system of the reactor design. Included are descriptions of the prestressed concrete reactor vessel that supports the magnets and contains the blanket and power conversion system components, the blanket module design, the blanket fuel design, and the power conversion system

Comparative studies of CERCER and CERMET fuels for EFIT from the viewpoint of core performance and safety

International Nuclear Information System (INIS)

Chen, X.N.; Rineiski, A.; Maschek, W.; Liu, P.; Boccaccini, C.M.; Sobolev, V.; Delage, F.; Rimpault, G.

2011-01-01

The European Facility for Industrial Transmutation (EFIT) has been developed within the 6. EU Framework by the EUROTRANS Program, aiming at a generic conceptual design of an accelerator driven transmuter. This paper deals with assessments of EFIT cores with CERCER and CERMET fuels from the viewpoint of core performance and safety. The conclusive remarks can be drawn as follows. Because of its much better thermal conductivity, the CERMET core can be designed by using thicker pins, so that it has the same or even better transmutation performance compared to the CERCER core. Both CERCER and CERMET fuels fulfill safety requirements. Moreover the CERMET fuel has higher fuel safety margins than the CERCER one. Preliminary analyses show that the CERMET total core power can be further increased by 50% at least without exceeding fuel and clad temperature limits. (authors)

The conceptualization model problem—surprise

Science.gov (United States)

Bredehoeft, John

2005-03-01

The foundation of model analysis is the conceptual model. Surprise is defined as new data that renders the prevailing conceptual model invalid; as defined here it represents a paradigm shift. Limited empirical data indicate that surprises occur in 20-30% of model analyses. These data suggest that groundwater analysts have difficulty selecting the appropriate conceptual model. There is no ready remedy to the conceptual model problem other than (1) to collect as much data as is feasible, using all applicable methods—a complementary data collection methodology can lead to new information that changes the prevailing conceptual model, and (2) for the analyst to remain open to the fact that the conceptual model can change dramatically as more information is collected. In the final analysis, the hydrogeologist makes a subjective decision on the appropriate conceptual model. The conceptualization problem does not render models unusable. The problem introduces an uncertainty that often is not widely recognized. Conceptual model uncertainty is exacerbated in making long-term predictions of system performance. C'est le modèle conceptuel qui se trouve à base d'une analyse sur un modèle. On considère comme une surprise lorsque le modèle est invalidé par des données nouvelles; dans les termes définis ici la surprise est équivalente à un change de paradigme. Des données empiriques limitées indiquent que les surprises apparaissent dans 20 à 30% des analyses effectuées sur les modèles. Ces données suggèrent que l'analyse des eaux souterraines présente des difficultés lorsqu'il s'agit de choisir le modèle conceptuel approprié. Il n'existe pas un autre remède au problème du modèle conceptuel que: (1) rassembler autant des données que possible en utilisant toutes les méthodes applicables—la méthode des données complémentaires peut conduire aux nouvelles informations qui vont changer le modèle conceptuel, et (2) l'analyste doit rester ouvert au fait

Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

Energy Technology Data Exchange (ETDEWEB)

Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

2006-07-01

Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

On the Impact of the Fuel Dissolution Rate Upon Near-Field Releases From Nuclear Waste Disposal

Directory of Open Access Journals (Sweden)

A Pereira

2016-09-01

Full Text Available Calculations of the impact of the dissolution of spent nuclear fuel on the release from a damaged canister in a KBS-3 repository are presented. The dissolution of the fuel matrix is a complex process and the dissolution rate is known to be one of the most important parameters in performance assessment models of the near-field of a geological repository. A variability study has been made to estimate the uncertainties associated with the process of fuel dissolution. The model considered in this work is a 3D model of a KBS-3 copper canister. The nuclide used in the calculations is Cs-135. Our results confirm that the fuel degradation rate is an important parameter, however there are considerable uncertainties associated with the data and the conceptual models. Consequently, in the interests of safety one should reduce, as far as possible, the uncertainties coupled to fuel degradation.

Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System

International Nuclear Information System (INIS)

1984-07-01

The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables

Safety methodology implementation in the conceptual design phase of a fusion reactor

International Nuclear Information System (INIS)

Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

2007-01-01

The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

Building a Conceptual Framework: Philosophy, Definitions, and Procedure

OpenAIRE

Yosef Jabareen

2009-01-01

In this paper the author proposes a new qualitative method for building conceptual frameworks for phenomena that are linked to multidisciplinary bodies of knowledge. First, he redefines the key terms of concept, conceptual framework, and conceptual framework analysis. Concept has some components that define it. A conceptual framework is defined as a network or a “plane” of linked concepts. Conceptual framework analysis offers a procedure of theorization for building conceptual frameworks base...

Conceptual design of tritium accountancy system for LLCB TBM

International Nuclear Information System (INIS)

Patel, Rudreksh; Sircar, Amit

2017-01-01

Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) will be tested in ITER for performance evaluation of high grade of heat extraction and tritium breeding. The bred tritium in the breeder materials is extracted and recovered by Tritium Extraction System (TES), whereas tritium permeated from breeder materials to helium coolants, viz., primary coolant and secondary coolant, is recovered by Coolant Purification System (CPS). This recovered tritium has to be accounted before transferring it to tritium plant (i.e., ITER inner fuel). This tritium accountancy is performed by Tritium Accountancy System (TAS). In addition to tritium accountancy, TAS also provides necessary data for the validation of design and modelling tools.In this work, we have presented conceptual design of TAS. It also describes operational philosophy, process parameters, process flow diagram, and interface details with ITER tritium plant. (author)

Nuclear data requirements for the ADS conceptual design EFIT: Uncertainty and sensitivity study

Energy Technology Data Exchange (ETDEWEB)

Garcia-Herranz, N., E-mail: nuria@din.upm.e [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (Spain); Cabellos, O. [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (Spain); Alvarez-Velarde, F. [CIEMAT (Spain); Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (Spain); Departamento de Ingenieria Energetica, UNED (Spain); Gonzalez-Romero, E.M. [CIEMAT (Spain); Juan, J. [Laboratorio de Estadistica, Universidad Politecnica de Madrid (Spain)

2010-11-15

In this paper, we assess the impact of activation cross-section uncertainties on relevant fuel cycle parameters for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) with a 'double strata' fuel cycle. Next, the nuclear data requirements are evaluated so that the parameters can meet the assigned design target accuracies. Different discharge burn-up levels are considered: a low burn-up, corresponding to the equilibrium cycle, and a high burn-up level, simulating the effects on the fuel of the multi-recycling scenario. In order to perform this study, we propose a methodology in two steps. Firstly, we compute the uncertainties on the system parameters by using a Monte Carlo simulation, as it is considered the most reliable approach to address this problem. Secondly, the analysis of the results is performed by a sensitivity technique, in order to identify the relevant reaction channels and prioritize the data improvement needs. Cross-section uncertainties are taken from the EAF-2007/UN library since it includes data for all the actinides potentially present in the irradiated fuel. Relevant uncertainties in some of the fuel cycle parameters have been obtained, and we conclude with recommendations for future nuclear data measurement programs, beyond the specific results obtained with the present nuclear data files and the limited available covariance information. A comparison with the uncertainty and accuracy analysis recently published by the WPEC-Subgroup26 of the OECD using BOLNA covariance matrices is performed. Despite the differences in the transmuter reactor used for the analysis, some conclusions obtained by Subgroup26 are qualitatively corroborated, and improvements for additional cross sections are suggested.

Multi-gate Pitot tube for the measurement of water flow in cooling systems; Tubo de Pitot multi-puertos para la medicion de flujo de agua en sistemas de enfriamiento

Energy Technology Data Exchange (ETDEWEB)

Figueroa Ibarra, Luis; Rodriguez Martinez, Jose Hugo; Santabarbara Botello, Marcelino [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)

2007-07-01

In this article the design of a novel measuring device of water flowing in large diameter pipes is described. A description is made of the instrument commonly used at present for this purpose (simple Pitot tube), whose design is based on the Standard CTI Code ATC-105 published by the Cooling Tower Institute (CTI ATC-105, 2000). The disadvantages of the simple Pitot tube in comparison with the new proposed device are emphasized (Multi-gate Pitot tube), of which its construction and design is described in detail. In the present article are also shown the results obtained of the circulation water flow measurements, performed in the cooling system of a fossil fuel power plant. These results are compared with the results of flow measurement obtained by means of the simulation of the thermodynamic cycle of the power station, using commercial software. [Spanish] En este articulo se describe el diseno de un novedoso medidor de flujo de agua que circula en tuberias de gran diametro. Se hace una descripcion del instrumento comunmente usado en la actualidad para dicho proposito (tubo de Pitot simple), cuyo diseno se basa en el estandar CTI Code ATC-105 publicado por el Cooling Tower Institute (CTI ATC-105, 2000). Se puntualizan las desventajas del tubo de Pitot simple en comparacion con el nuevo dispositivo propuesto (tubo de Pitot Multi-puertos), del cual se describe detalladamente su construccion y diseno. En el presente articulo tambien se muestran los resultados obtenidos de las mediciones de flujo de agua de circulacion, realizadas en el sistema de enfriamiento de una central termoelectrica. Dichos resultados son comparados con los resultados de flujo obtenidos mediante simulacion del ciclo termodinamico de la central, usando un software comercial.

Design of fuel element for RA10

International Nuclear Information System (INIS)

Estevez, Esteban A.; Markiewicz, Mario; Gerding, Roberto

2012-01-01

The RA-10 reactor is an open pool multipurpose reactor. It is intended for radioisotopes production, fuel irradiation and use of neutron beam experiments. The nominal configuration core consists of 19 fuel elements (FE) and 6 in-core irradiation positions. With regard to the FE, although both conceptual design and manufacturing technology are similar to the already developed and qualified by CNEA (MTR fuel flat plate), the conditions imposed by the new reactor on FE's are more demanding that previous supplies. Here it should be mentioned the magnitude of the hydrodynamic forces acting on the FE caused by coolant flow through the core (upward) and mainly by the high coolant velocity between fuel plates (greater than 5 times than those currently in operation). Moreover, the high power density results in higher heat flux in fuel plates and greater temperature gradient. As a result of these increased demands present during irradiation, and in order to maintain a high level of reliability, it is necessary carry out some modifications in the mechanical design of the FE (with respect to the so-called ECBE design or s tandard ) . Design verification is performed through analytical and code calculations, and hydrodynamic tests on a full-scale prototype. This article describes the design of the FE for RA 10 reactor, with special emphasis on those aspects that represent innovations in the traditional design (ECBE). It also presents the functional requirements, design criteria and design limits established according to the reactor operational states (author)

Design and analysis of an intelligent public FV lighting system; Diseno y analisis de un alumbrado publico FV inteligente

Energy Technology Data Exchange (ETDEWEB)

Hanessian D, Ana V.; Gordon, Manuel [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Mexico, D.F. (Mexico)

2009-07-01

In the Mexico's National Energy Balance of 2008, it is considered that of the total of the electrical power consumption in our country, 18% is dedicated to lighting. Conscious of the necessity of saving energy in public lighting in this article is presented the design, construction and analysis of the power consumption of a public light fed with electricity of photovoltaic cells and the control of intensity on the light in inverse way of the natural light. A lamp constructed based in light emitting diodes (LEDs) is used. This has the quality of consuming very little energy that could be provided by the sun and be stored to use it at night. With this system, proven at scale, energy savings are obtained superior to 50% of the conventional one and, in relation to the commercial photovoltaic (FV) luminaries up to 30%. [Spanish] Del Balance Nacional de Energia, de Mexico, de 2008, se considera que del total del consumo de energia electrica en nuestro pais, el 18% esta dedicado a la iluminacion. Conscientes de la necesidad de ahorrar energia en alumbrado publico, en este articulo se presenta el diseno de construccion y analisis del consumo energetico de una luminaria publica alimentada con electricidad de celdas fotovoltaicas y el control de intensidad de la luz de manera inversa a la luz natural. Se utiliza un foco construido a base a los diodos emisores de luz (LEDs, por sus siglas en ingles). Estos tienen la cualidad de consumir muy poca energia que podra ser suministrada por el sol y almacenada para utilizarla en la noche. Con este sistema, probado a escala, se logran ahorros de energia superiores al 50% del convencional y, en relacion a las luminarias fotovoltaicas (FV) comerciales, hasa el 30%.

Kuhn and conceptual change: on the analogy between conceptual changes in science and children

Science.gov (United States)

Greiffenhagen, Christian; Sherman, Wendy

2008-01-01

This article argues that the analogy between conceptual changes in the history of science and conceptual changes in the development of young children is problematic. We show that the notions of ‘conceptual change’ in Kuhn and Piaget’s projects, the two thinkers whose work is most commonly drawn upon to support this analogy, are not compatible in the sense usually claimed. We contend that Kuhn’s work pertains not so much to the psychology of individual scientists, but to the way philosophers and historians should describe developments in communities of scientists. Furthermore, we argue that the analogy is based on a misunderstanding of the nature of science and the relation between science and common sense. The distinctiveness of the two notions of conceptual change has implications for science education research, since it raises serious questions about the relevance of Kuhn’s remarks for the study of pedagogical issues.

Aviation Management Perception of Biofuel as an Alternative Fuel Source

Science.gov (United States)

Marticek, Michael

The purpose of this phenomenological study was to explore lived experiences and perceptions from a population of 75 aviation managers in various locations in Pennsylvania about the use of aviation biofuel and how it will impact the aviation industry. The primary research question for this study focused on the impact of biofuel on the airline industry and how management believes biofuel can contribute to the reduction of fossil fuel. Grounded in the conceptual framework of sustainability, interview data collected from 27 airline and fueling leaders were analyzed for like terms, coded, and reduced to 3 themes. Data were organized and prioritized based on frequency of mention. The findings represented themes of (a) flight planning tools, (b) production, and (c) costs that are associated with aviation fuel. The results confirmed findings addressed in the literature review, specifically that aviation biofuel will transform the airline industry through lower cost and production. These findings have broad applicability for all management personnel in the aviation industry. Implications for social change and improved business environments could be realized with a cleaner environment, reduced fuel emissions, and improved air quality.

Conceptual design for accelerator-driven sodium-cooled sub-critical transmutation reactors using scale laws

Energy Technology Data Exchange (ETDEWEB)

Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates the reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scaled and verified through the methodology in this paper, which is referred to Advanced Liquid Metal Reactor (ALMR). A Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the integrated code system. Integrated Code System (ICS) consists of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related regions are analyzed once-through Results of conceptual design are attached in this paper. 5 refs., 4 figs., 1 tab. (Author)

Conceptualizing Telehealth in Nursing Practice: Advancing a Conceptual Model to Fill a Virtual Gap.