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Sample records for gen iv vhtr

  1. Gen IV Materials Handbook Implementation Plan

    International Nuclear Information System (INIS)

    Rittenhouse, P.; Ren, W.

    2005-01-01

    A Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Handbook will be responsive to the needs expressed by all of the principal government, national laboratory, and private company stakeholders of Gen IV Reactor Systems. The Gen IV Materials Handbook Implementation Plan provided here addresses the purpose, rationale, attributes, and benefits of the Handbook and will detail its content, format, quality assurance, applicability, and access. Structural materials, both metallic and ceramic, for all Gen IV reactor types currently supported by the Department of Energy (DOE) will be included in the Gen IV Materials Handbook. However, initial emphasis will be on materials for the Very High Temperature Reactor (VHTR). Descriptive information (e.g., chemical composition and applicable technical specifications and codes) will be provided for each material along with an extensive presentation of mechanical and physical property data including consideration of temperature, irradiation, environment, etc. effects on properties. Access to the Gen IV Materials Handbook will be internet-based with appropriate levels of control. Information and data in the Handbook will be configured to allow search by material classes, specific materials, specific information or property class, specific property, data parameters, and individual data points identified with materials parameters, test conditions, and data source. Details on all of these as well as proposed applicability and consideration of data quality classes are provided in the Implementation Plan. Website development for the Handbook is divided into six phases including (1) detailed product analysis and specification, (2) simulation and design, (3) implementation and testing, (4) product release, (5) project/product evaluation, and (6) product

  2. Gen IV. Technical and economical aspects

    International Nuclear Information System (INIS)

    Kaluzny, Y.; Legee, F.

    2010-01-01

    In this presentation author deals with development of nuclear reactor type of Generation IV. He concluded that: - Nuclear energy is competitive with regards to the other generation sources; Its competitiveness also increases with CO 2 cost. Considering the nuclear cost breakdown of LWR reactors, it turns out that the uranium is currently not in the range of a threshold for FBR deployment; - The global balance of uranium supply and demand and also innovation required to fulfil GEN IV objectives would probably imply the emergence of fast reactor competitiveness after the turn of the mid-century; - We shall need fast reactors in the coming decade.

  3. Gen IV Materials Handbook Functionalities and Operation

    International Nuclear Information System (INIS)

    Ren, Weiju

    2009-01-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  4. Gen IV Materials Handbook Functionalities and Operation

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL

    2009-12-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  5. Euratom research and training in generation IV systems with emphasis on V/HTR

    International Nuclear Information System (INIS)

    Goethem, G. van; Manolatos, P.; Fuetterer, M.

    2006-01-01

    In this overview paper, the following questions are addressed: (1) What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2) What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle: Education, Research, and Innovation') respond to the S/T challenges: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is principally under the responsibility of the 2 Directorates Generals (DG) DG Research (RTD, located in Brussels), which implements and manages the programme of 'indirect actions', and the DG Joint Research Centre (JRC, headquarters in Brussels and 7 scientific institutes in 5 Member States) which carries out 'direct actions' in their own laboratories. In this HTR-2006 introductory paper, the emphasis is on the indirect and direct actions of the 6 th Euratom research framework programme 2003-2006, FP-6, with special emphasis on V/HTR Generation IV research. (orig.)

  6. Improvement of Steam Generator Reliability for GEN-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

    2005-11-15

    The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator.

  7. Improvement of Steam Generator Reliability for GEN-IV SFR

    International Nuclear Information System (INIS)

    Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

    2005-11-01

    The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator

  8. New Materials for NGNP/Gen IV

    International Nuclear Information System (INIS)

    Swindeman, Robert W.; Marriott, Douglas L.

    2009-01-01

    The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.

  9. ASN’s actions in GEN IV reactors and Sodium Fast Reactors (SFR)

    International Nuclear Information System (INIS)

    Belot, Clotilde

    2013-01-01

    The ASN is involved in 3 actions concerning GEN IV: • Overview of nuclear reactor GEN IV systems; • Specific analysis about transmutation; • Prototype reactor ASTRID (SFR). Furthermore theses actions are in the beginning (no conclusions or results available)

  10. Market share scenarios for Gen-DIII and gen-IV reactors in Europe

    International Nuclear Information System (INIS)

    Roelofs, F.; Heek, A. V.; Durpel, L. V. D.

    2008-01-01

    Nuclear energy is back on the agenda worldwide in order to meet growing energy demand and especially the growth in electricity demand. Many objectives direct to an increased use of nuclear energy, i.e. minimising energy costs, reducing climate change effects and others. In the light of the potential renewed growth of nuclear energy, the public demands a clear view on what nuclear energy may contribute towards meeting these objectives and especially how nuclear energy may address some socio-political obstructions with respect to economics, radioactive waste, safety and proliferation of fissile materials. To address these questions, the future nuclear reactor park mix in Europe has been analysed applying an integrated dynamic process modelling technique. Various market share scenarios for nuclear energy are derived including sub-variants with regard to the intra-nuclear options. In the analyses, it is assumed that different types of new reactors may be built, taking into account the introduction date of considered Gen-Ill (i.e. EPR) and Gen-IV (i.e. SCWR, HTR, FR) reactors, and the economic evaluation of the complete fuel cycle. The assessment was undertaken using the DANESS code (Dynamic Analysis of Nuclear Energy System Strategies). The analyses show that given the considered realistic nuclear energy demand and given a limited number of available Gen-III and Gen-IV reactor types, the future European nuclear park will exist of combinations of Gen-III and Gen-IV reactors. This mix will always consist of a set of reactor types each having its specific strengths. The analyses also highlight the triggers influencing the choice between different nuclear energy deployment scenarios. (authors)

  11. GEN IV reactors: Where we are, where we should go

    International Nuclear Information System (INIS)

    Locatelli, G.; Mancini, M.; Todeschini, N.

    2012-01-01

    GEN IV power plants represent the mid-long term option of the nuclear sector. International literature proposes many papers and reports dealing with these reactors, but there is an evident difference of type and shape of information making impossible each kind of detailed comparison. Moreover, authors are often strongly involved in some particular design; this creates many difficulties in their super-partes position. Therefore it is necessary to put order in the most relevant information to understand strengths and weaknesses of each design and derive an overview useful for technicians and policy makers. This paper presents the state-of the art for GEN IV nuclear reactors providing a comprehensive literature review of the different designs with a relate taxonomy. It presents the more relevant references, data, advantages, disadvantages and barriers to the adoptions. In order to promote an efficient and wide adoption of GEN IV reactors the paper provides the pre-conditions that must be accomplished, enabling factors promoting the implementation and barriers limiting the extent and intensity of its implementation. It concludes outlying the state of the art of the most important R and D areas and the future achievements that must be accomplished for a wide adoption of these technologies. (authors)

  12. GIF (Gen-IV International Forum) Symposium 2009. Proceedings

    International Nuclear Information System (INIS)

    2009-01-01

    The objective of this symposium is to give a well documented state of the art of the initiative and to report and discuss the most significant technical progress and evolution in the different areas during these last ten years. Another significant objective is to provide a forum for an open and hopefully lively discussion of the perspectives, priorities and challenges for the next few years, accounting for a rapidly evolving environment. The symposium has been organized into three sessions that have dealt with the following issues: -) Generation IV International Forum (GIF): 10 years of achievements and the path forward, -) Methodology Overviews and Focus on Applications, -) Very High Temperature Reactor (VHTR), -) Gas-cooled Fast Reactor (GFR), -) Super-Critical Water-cooled Reactor (SCWR), -) Lead-cooled Fast Reactor (LFR), -) Molten Salt Reactor (MSR), -) Sodium-cooled Fast Reactor (SFR), -) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and its potential synergy with GIF, and -) GIF priority objectives for the next 5 years

  13. Gen IV International Forum - GIF, 2010 Annual Report

    International Nuclear Information System (INIS)

    Anon.

    2011-01-01

    The Generation IV International Forum (GIF), created in 2000 to foster international collaboration at a detailed level of actual R and D, is a cooperative international endeavor, organized to develop the research necessary to test the feasibility and performance capabilities of fourth generation nuclear systems, with the goal of making such systems deployable in large numbers around 2030. Since its beginning, GIF members stated the following goals for the fourth generation of nuclear power plants when compared to previous generations: a) improve sustainability (including effective fuel utilization and minimization of waste); b) improve economics (competitiveness with respect to other energy sources); c) improve safety and reliability (e.g. no need for offsite emergency response); and d) improve proliferation resistance and physical protection. After an in-depth analysis of the different available concepts, whatever their level of development, the Forum selected six concepts as the most promising, and decided to focus R and D on these systems: - the very-high-temperature reactor (VHTR); - the sodium-cooled fast reactor (SFR); - the supercritical-water-cooled reactor (SCWR); - the gas-cooled fast reactor (GFR); - the lead-cooled fast reactor (LFR); - the molten salt reactor (MSR). Active members of the GIF are Canada, Euratom, France, Japan, People's Republic of China, Republic of Korea, Republic of South Africa, Russian Federation, Switzerland and the United States. Altogether, they represent around 90% of the world installed nuclear capacity for producing electricity, and all key technology holders. The forum is led by the policy group, where all members are represented, and currently chaired by Japan since December 2009, assisted by vice-chairs from France and United States. The year 2010 has seen some important achievements and decisions regarding these six systems. For example, two sodium-cooled fast reactors (re)started this year: Monju in Japan restarted after

  14. Reactor physics challenges in GEN-IV reactor design

    International Nuclear Information System (INIS)

    Driscoll, Michael K.; Hejzlar, Pavel

    2005-01-01

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

  15. Reactor physics challenges in GEN-IV reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

    2005-02-15

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

  16. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors

    International Nuclear Information System (INIS)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-01-01

    Many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important criterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals

  17. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-04-23

    many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

  18. JRC-IE's research of safety of Gen IV systems

    International Nuclear Information System (INIS)

    Tsige-Tamirat, H.; Ranguelova, V.; Feutterer, M.; Ammirabile, L.; Carlsson, J.; D'Agata, E.; Laurie, M.; Magallon, D.

    2010-01-01

    The Institute for Energy (IE), one of the seven scientific Institutes of the Joint Research Centre (JRC) of the European Commission, has the mission to provide scientific and technical support for the conception, development, implementation and monitoring of community policies related to energy. To accomplish its mission, IE performs research in the areas of renewable energies, safety and sustainability of nuclear energy for current and future reactor systems, energy technic/economic assessment, and security of energy supply. The Generation IV International Forum (GIF) is a cooperative international endeavour organized to carry out R and D needed to establish the feasibility and performance capabilities of the next generation nuclear energy systems and support the progress towards their realization. The EU, represented by EURATOM and with the JRC as implementing agent, is working together with other GIF partners to perform pre-competitive R and D on key technologies to be implemented in future nuclear systems. IE is engaged in experimental research, simulation and modeling, scientific, feasibility and engineering studies on innovative nuclear reactor systems needed to support the EURATOM contribution to GEN IV initiative, in particular in assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions and knowledge management and preservation. IE's research activities on Generation IV reactor systems are focused on the assessment of the potential of such systems to meet long term EU energy needs with respect to economical advantages, enhanced safety, sustainability, and proliferation resistance. IE participates in international collaborations and has bilateral research cooperation both with European and non-European partners. This paper gives an overview of IE's current research activities on the Gen IV reactor systems related to safety. (authors)

  19. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

    International Nuclear Information System (INIS)

    Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  20. Developing new nuclear curricula for GEN IV needs

    International Nuclear Information System (INIS)

    Ghitescu, P.; Pavel, G.L.

    2014-01-01

    States who wish to start and develop a nuclear program must take into consideration a strong proven strategy for developing a sustainable program. A complete nuclear research program must include: a good national strategy and support on the topic; strong research laboratories supported by good personnel; education component to provide sustainable and qualified workforce; national/international interest from stakeholders and governments and a well informed society. New demonstrators are foreseen for the next period to be built in Europe and skilled supporting personnel is strongly needed. Current situation in nuclear higher education with perspective will be analysed. EURATOM strongly supports development of multidisciplinary co-operational projects in order to built such novel initiatives. An example of such program supported by European Commission, ARCADIA, will be given. The project is based on the cooperation of a large number of participants all over Europe and the main purpose is to develop a road-map for Gen IV reactor. (authors)

  1. Environmental sensitivity studies for Gen-IV roadmap DUPIC scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2004-03-01

    The environmental effect of the DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel cycle, which is considered as one of the partial recycle scenario in Gen-IV roadmap, has been analyzed by using the dynamic analysis method. Through the parametric calculations for the DUPIC fuel cycle deployment time and the fraction of the DUPIC reactors, the environmental effects of the fuel cycle for important parameters such as the amount of spent fuel and the combined amounts of plutonium and minor actinides were estimated and compared to those of the once-through LWR fuel cycle. The results of the sensitivity calculations showed that an early deployment of the DUPIC fuel cycle with a high DUPIC reactor fraction can reduce the accumulation of spent fuel by up to 40%. More important is the associated reduction in the combined amount of plutonium and minor actinides, which may reduce the key repository parameter (long term decay heat). Therefore it is expected that favorable environmental effects will be the outcome of the implementation of the DUPIC fuel cycle

  2. Emissivity of Candidate Materials for VHTR Applicationbs: Role of Oxidation and Surface Modification Treatments

    International Nuclear Information System (INIS)

    Sridharan, Kumar; Allen, Todd; Anderson, Mark; Cao, Guoping; Kulcinski, Gerald

    2011-01-01

    The Generation IV (GEN IV) Nuclear Energy Systems Initiative was instituted by the Department of Energy (DOE) with the goal of researching and developing technologies and materials necessary for various types of future reactors. These GEN IV reactors will employ advanced fuel cycles, passive safety systems, and other innovative systems, leading to significant differences between these future reactors and current water-cooled reactors. The leading candidate for the Next Generation Nuclear Plant (NGNP) to be built at Idaho National Lab (INL) in the United States is the Very High Temperature Reactor (VHTR). Due to the high operating temperatures of the VHTR, the Reactor Pressure Vessel (RPV) will partially rely on heat transfer by radiation for cooling. Heat expulsion by radiation will become all the more important during high temperature excursions during off-normal accident scenarios. Radiant power is dictated by emissivity, a material property. The NGNP Materials Research and Development Program Plan (1) has identified emissivity and the effects of high temperature oxide formation on emissivity as an area of research towards the development of the VHTR.

  3. Safety Design Criteria (SDC) for Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Nakai, Ryodai

    2013-01-01

    SDC Development Background & Objectives: • Safety Design Criteria (SDC) Development for Gen-IV SFR: – Proposed at the GIF Policy Group (PG) meeting in October 2010 –SDC “harmonization” is increasingly important for: • Realization of enhanced safety designs meeting to Gen-IV safety goals and safety approach common to SFR systems; • Preparation for the forthcoming licensing in the near future; • Because Gen-IV SFR are progressing into conceptual design stage. • The SDC is the Reference criteria: – Of the designs of safety-related Structures, Systems & Components that are specific to the SFR system; – For clarifying the requisites systematically & comprehensively; – When the technology developers apply the basic safety approach and use the codes & standards for conceptual design of the Gen-IV SFR system

  4. European project SARGEN IV: safety approach and assessment of GEN IV reactors

    International Nuclear Information System (INIS)

    Ammirabile, L.

    2013-01-01

    • SARGEN I V has elaborated a proposal for the harmonization of safety assessment practices for GEN IV NPP. • An overall reinforcement of DiD is expected for GEN I V NPP, including improved independence between all levels of DiD. • An inherent approach should reinforce the fulfillment of fundamental safety functions e.g. the consequences for some situations should be reduced and the grace periods should be extended. For the same reason, the use of passive systems can be envisaged. • The need of complementary and integrated deterministic and probabilistic approaches is reiterated. • Methodologies: Some of them are not yet applied. • Assessment of hazards would be a challenging aspect of next generation of NPP safety assessment and should be improved, which is confirmed by the first insights of Fukushima Daiichi TEPCO reactors accidents. • Provisions to cope with extreme events notably to improve the grace period before cliff-edge effects and thus allowing back-up measures to be implemented have to be defined and should be considered as hardened equipments

  5. Validation Studies for Numerical Simulations of Flow Phenomena Expected in the Lower Plenum of a Prismatic VHTR Reference Design

    International Nuclear Information System (INIS)

    Richard W. Johnson

    2005-01-01

    The final design of the very high temperature reactor (VHTR) of the fourth generation of nuclear power plants (Gen IV) has not yet been established. The VHTR may be either a prismatic (block) or pebble bed type. It may be either gas-cooled or cooled with an as yet unspecified molten salt. However, a conceptual design of a gas-cooled VHTR, based on the General Atomics GT-MHR, does exist and is called the prismatic VHTR reference design, MacDonald et al [2003], General Atomics [1996]. The present validation studies are based on the prismatic VHTR reference design. In the prismatic VHTR reference design, the flow in the lower plenum will be introduced by dozens of turbulent jets issuing into a large crossflow that must negotiate dozens of cylindrical support columns as it flows toward the exit duct of the reactor vessel. The jets will not all be at the same temperature due to the radial variation of power density expected in the core. However, it is important that the coolant be well mixed when it enters the power conversion unit to ensure proper operation and long life of the power conversion machinery. Hence, it is deemed important to be able to accurately model the flow and mixing of the variable temperature coolant in the lower plenum and exit duct. Accurate flow modeling involves determining modeling strategies including the fineness of the grid needed, iterative convergence tolerance, numerical discretization method used, whether the flow is steady or unsteady, and the turbulence model and wall treatment employed. It also involves validation of the computer code and turbulence model against a series of separate and combined flow phenomena and selection of the data used for the validation. The present report describes progress made to date for the task entitled ''CFD software validation of jets in crossflow'' which was designed to investigate the issues pertaining to the validation process

  6. Development of Basic Key Technologies for Gen IV SFR

    International Nuclear Information System (INIS)

    Han, Do Hee; Kim, Young In; Won, Byung Chool

    2008-11-01

    Technical specifications such as power capacity, type of core, clad alloy, clad barrier material, number of loops, type of SG tube have been evaluated and a optimal design concept has been identified to satisfy the technology goals of Generation IV nuclear systems. The concept for breakeven design is featured by the heat capacity of 1,200 MWe, enrichment-separated core, 2-loop, double-walled SG tube, and a long-life sensor system for in-service inspection

  7. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations. Published on September 24, 2012

    International Nuclear Information System (INIS)

    Couturier, Jean; Bruna, Giovanni; Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Hache, Georges

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  8. Development of basic key technologies for Gen IV SFR

    International Nuclear Information System (INIS)

    Kim, Yeongil; Kim, Sungoh; Choi, Sukgi

    2012-04-01

    The advanced concepts, for the breakeven reactor(1,200MWe) and TRU burner(600MWe), were defined to satisfy the technology goals of Generation IV nuclear systems. Based on the advanced design concepts, a conceptual design of the demonstration SFR has been developed using the available licensing technology. The conceptual core design has been developed for the TRU burner in which an initial core is fueled with less than 20wt% enriched U235, and finally transformed to a self-recycled TRU core. The passive decay heat removal circuit ensuring reactor safety even in case of loss of emergency power has been developed and minimization of a reactor vessel and simplification of reactor internals have been conducted in the conceptual design. For development of advanced technologies, control logics for various power levels and the optimal design concept of heat exchanger applicable to supercritical CO 2 Brayton cycle as an energy conversion system was developed. A novel under-sodium waveguide sensor and a prototype under-sodium inspection system have been developed for under-sodium viewing of in-vessel structures submerged in an opaque liquid sodium. The fabrication technology of fuel slugs using the advanced fuel slug casting system was developed, and U-Zr alloy fuel rods were fabricated and examined. And a HT 9 cladding tube was manufactured using the developed cladding tube fabrication technology. For development of basic technologies, the cross section adjustment code ATCROSS and the MATRA-LMR code with HCFs have been developed to reduce core design uncertainties. The SIE ASME-NH computer program to evaluate high temperature structural design for 60 years design life, and the safety analysis code MARS-LMR with thermal-hydraulic and reactivity feedback models have been developed and validated. In addition, the sodium impurity measurement and control technology, the sodium water reaction event propagation model to predict the sodium leak propagation in a steam generator, and

  9. Safeguards Licensing Aspects of a Future Gen IV Test Facility - a Case Study

    International Nuclear Information System (INIS)

    Lindell, M. Aberg; Grape, S.; Hakansson, A.; Svaerd, S. Jacobsson

    2010-01-01

    The scope of this study covers safeguards licensing aspects of a possible future Gen IV demonstration facility. As a basis for the investigation, the facility was assumed to be located in Sweden, comprising a lead-cooled fast reactor and a reprocessing plant with fuel fabrication. The aim has been to identify safeguards requirements that may be set by the IAEA and the Swedish Radiation Safety Authority, and also to suggest how the safeguards system could be implemented in practice. The changed usage and handling of nuclear fuel, as compared to that of today, has been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. This work is part of GENIUS, the Swedish Gen IV research and development programme, which emphasizes lead-cooled fast reactors. (author)

  10. Thermal stability study for candidate stainless steels of GEN IV reactors

    International Nuclear Information System (INIS)

    Simeg Veternikova, J.; Degmova, J.; Pekarcikova, M.; Simko, F.; Petriska, M.; Skarba, M.; Mikula, P.; Pupala, M.

    2016-01-01

    Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  11. Thermal stability study for candidate stainless steels of GEN IV reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simeg Veternikova, J., E-mail: jana.veternikova@stuba.sk [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Degmova, J. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pekarcikova, M. [Institute of Materials Science, Faculty of Materials Science and Technology, Slovak University of Technology, Paulinska 16, 917 24 Trnava (Slovakia); Simko, F. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia); Petriska, M. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Skarba, M. [Slovak University of Technology, Vazovova 5, 812 43 Bratislava (Slovakia); Mikula, P. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pupala, M. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia)

    2016-11-30

    Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  12. Fuel research for subcritical and critical GEN-IV systems cooled by heavy liquid metal

    International Nuclear Information System (INIS)

    Sobolev, V.; Verwerft, M.

    2009-01-01

    The participation of the Belgian Nuclear Research Centre SCK-CEN in the worldwide GEN-IV research can be considered as an opportunity. Today's GEN-IV research at SCK-CEN is mainly driven by the interests of the project MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications). The main goal of this project is to build at SCK-CEN in Mol a new generation fast spectrum, subcritical, research and materials testing reactor MYRRHA driven by a high-energy proton accelerator. This GEN-IV MTR is cooled by heavy liquid metal (Pb-Bi) and will be used for the ADS concept demonstration, testing and qualification of new fuels, transmutation targets and innovative materials. On the European scale, MYRRHA is integrated in the Euratom FP6 Integrated Project (IP) EUROTRANS (EUROpean research programme for TRANSmutation of high level nuclear waste in an accelerator driven system), as the small-scale experimental machine for transmutation demonstration called XT-ADS. Last but not least, this experimental facility will also demonstrate the technological feasibility of the LFR (Lead-cooled Fast Reactor) GEN-IV concept; in EU the LFR design studies are performed in the framework of the Euratom FP6 ELSY (European Lead-cooled SYstem) project, where SCK-CEN is a partner. Among the research needed to ensure a safe and reliable operation of the MYRRHA/XT ADS reactor, the development and qualification of fuel and cladding materials have been recognized as one of the main key issues to be addressed

  13. Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

    Science.gov (United States)

    Murty, K. L.; Charit, I.

    2008-12-01

    Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

  14. Gen IV Materials Handbook Functionalities and Operation (2B) Handbook Version 2.0

    International Nuclear Information System (INIS)

    Ren, Weiju

    2011-01-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  15. Gen IV Materials Handbook Functionalities and Operation (4A) Handbook Version 4.0

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL

    2013-09-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  16. Gen IV Materials Handbook Functionalities and Operation (2B) Handbook Version 2.0

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL

    2011-08-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  17. The status of proliferation resistance evaluation methodology development in GEN IV international forum

    International Nuclear Information System (INIS)

    Inoue, Naoko; Kawakubo, Yoko; Seya, Michio; Suzuki, Mitsutoshi; Kuno, Yusuke; Senzaki, Masao

    2010-01-01

    The Generation IV Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PR and PP WG) was established in December 2002 in order to develop the PR and PP evaluation methodology for GEN IV nuclear energy systems. The methodology has been studied and established by international consensus. The PR and PP WG activities include development of the measures and metrics; establishment of the framework of PR and PP evaluation, the demonstration study using Example Sodium Fast Reactor (ESFR), which included the development of three evaluation approaches; the Case Study using ESFR and four kinds of threat scenarios; the joint study with GIF System Steering Committees (SSCs) of the six reactor design concepts; and the harmonization study with the IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper reviews the status of GIF PR and PP studies and identifies the challenges and directions for applying the methodology to evaluate future nuclear energy systems in Japan. (author)

  18. Study on high temperature design methodology of heat-resistant materials for GEN-IV systems

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Seog; Kim, D. W.; Kim, S. H.; Kim, W. G.; Kim, J. H.; Park, D. G.; Yoon, J. H.; Lee, H. Y.; Hing, J. H

    2005-08-15

    Analysis of the existing high temperature design and assessment codes such as US(ASME-NH,Draft Code Case for Alloy 617), France(RCC-MR), UK(R5), Japan(BDS/DDS/FDS) for Gen IV reactor structure has been carried out. In addition the scope and fields for research and development is needed in the future have been defined. For assessing the high temperature creep cracks, time dependent fracture mechanics (TDFM) parameters of the C and Ct were analyzed. The creep propagation data were obtained from the creep crack growth tests for type 316LN stainless steels, and creep crack growth testing machine for Gen-IV system up to 950 .deg. C was set up. Damage mechanism and causes for creep-fatigue were investigated. The difference between prediction creep-fatigue life and experimental life were investigated. Material properties for analysis creep-fatigue damage were recommended. The assessment procedure (Draft) on creep-fatigue crack initiation has been developed based on the technical appendix A16 of French RCC-MR code. Ultrasonic wave signal against creep ruptured specimens of type 316LN stainless steel was obtained. It was identified that creep damage can be evaluated by ultrasonic method. The NDT techniques evaluated include Barkhausen noise, magnetic hysteresis parameters, positron annihilation, X-ray diffraction and small angle neutron scattering. Experimental procedure and evaluation method of material integrity were developed through the fracture toughness test of Cr-Mo steel.

  19. Analysis of Creep Crack Growth Behavior of Alloy 617 for Use in a VHTR System

    International Nuclear Information System (INIS)

    Kim, Woo-Gon; Kim, Min-Hwan; Park, Jae-Young; Ekaputra, I. M. W.; Kim, Seon-Jin

    2015-01-01

    Alloy 617 is a major candidate material for the IHX component. The design of the component, which will operate well into the creep range, will require a good understanding of creep crack growth deformation. Efforts are now being undertaken in the Gen-IV program to provide data needed for the design and licensing of the nuclear plants, and with this goal in mind, to meet the needs of the conceptual designers of the VHTR system, 'Gen-IV Materials Handbook' is being established through an international collaboration program of GIF (Gen-IV Forum) countries. To logically obtain the B and q values in the CCGR equation, three methods in terms of LSFM, MVM, and PDM were adopted. The PDM was most useful. Both the B and q coefficients followed a lognormal distribution. Using a lognormal distribution in the PDM, a number of random variables were generated by Monte Carlo Simulation, and the CCGR lines could be successfully predicted from the viewpoint of reliability

  20. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  1. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR and PP)

    International Nuclear Information System (INIS)

    Moses, David Lewis

    2011-01-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR and PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR and PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR and PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR and PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet

  2. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  3. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  4. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  5. Environmental sensitivity studies for Gen-IV roadmap fast reactor scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2004-03-01

    The environmental effect of the self-sufficient fast reactor scenario, which is considered as one of the full fissile plutonium and transuranic recycle scenario in Gen-IV roadmap, has been analyzed by using the dynamic analysis method. Through the parametric calculations for the fast reactor deployment time and capacity, the environmental effects of the fuel cycle for important parameters such as the amount of spent fuel and the combined amounts of plutonium and minor actinides were estimated and compared to those of the once-through LWR fuel cycle. The results of the sensitivity calculations showed that an early deployment of the fast reactor with a high capacity can reduce the accumulation of spent fuel by up to 37%. Furthermore, the recycling of plutonium and minor actinides can reduce the key repository parameter (long term decay heat). Therefore the favorable environmental effects can be expected with the implementation of the symbiotic fast reactor scenario

  6. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  7. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Tsige-Tamirat, H.; Ammirabile, L.; D' Agata, E.; Fuetterer, M.; Ranguelova, V. [European Commission, Joint Research Centre, Institute for Energy, Westerduinweg 3, 1755LE Petten (Netherlands)

    2010-07-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  8. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    International Nuclear Information System (INIS)

    Tsige-Tamirat, H.; Ammirabile, L.; D'Agata, E.; Fuetterer, M.; Ranguelova, V.

    2010-01-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  9. Feasibility study on the application of carbide (ZrC, SiC) for VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ji Yeon; Kim, Weon Ju; Jung, Choong Hwan; Ryu, Woo Seog; Kim, Si Hyeong; Jang, Moon Hee; Lee, Young Woo

    2006-08-15

    A feasibility study on the coating process of ZrC for the TRISO nuclear fuel and applications of SiC as high temperature materials for the core components has performed to develop the fabrication process for the advanced ZrC TRISO fuels and the high temperature structural components for VHTR, respectively. In the case of ZrC coating, studies were focused on the comparisons of the developed coating processes for screening of our technology, the evaluations of the reactions parameters for a ZrC deposition by the thermodynamic calculations and the preliminary coating experiments by the chloride process. With relate to SiC ceramics, our interesting items are as followings; an analysis of applications and specifications of the SiC components and collections of the SiC properties and establishments of data base. For these purposes, applications of SiC ceramics for the GEN-IV related components as well as the fusion reactor related ones were reviewed. Additionally, the on-going activities with related to the ZrC clad and the SiC composites discussed in the VHTR GIF-PMB, were reviewed to make the further research plans at the section 1 in chapter 3.

  10. Strategy of VHTR Realization

    International Nuclear Information System (INIS)

    Chang, Jonghwa

    2015-01-01

    High temperature gas cooled reactor has been developed since 1956. Fundamental idea of a gas cooled reactor is to achieve high temperature which is suitable for high efficiency application such as electricity generation. The core is composed of ceramics, graphite blocks which are mechanical stable up to very high temperature. Fuel is ceramics, TRISO ( tri-isotropic coated micro particle) whose dense coating layers work as small radioactivity containment. Coolant is inert gas, helium, which is stable chemically, neutronically, and thermal hydraulically. Several test reactors such as DRE, PB-1, FSV, AVR, THTR, HTTR, HTR-10 were built and demonstrated their safety. Large GA-HTR, RSA-PBMR projects are canceled and US-NGNP project is idling. Only Chinese HTR-PM demonstrator is under construction. HTGR has long history of development. For realization and market penetration, VHTR community should look at niche market such as carbon free energy supply to industry complex, electric power for small grid, carbon free hydrogen production, power source for space colony. Technology Readiness Level should be advanced to get proper investment from industry. For this, cooperation between international R and D institutions is required. Clearly divided role between universities, research institutions, and industries will reduce complication and shorten VHTR realization day

  11. Strategy of VHTR Realization

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    High temperature gas cooled reactor has been developed since 1956. Fundamental idea of a gas cooled reactor is to achieve high temperature which is suitable for high efficiency application such as electricity generation. The core is composed of ceramics, graphite blocks which are mechanical stable up to very high temperature. Fuel is ceramics, TRISO ( tri-isotropic coated micro particle) whose dense coating layers work as small radioactivity containment. Coolant is inert gas, helium, which is stable chemically, neutronically, and thermal hydraulically. Several test reactors such as DRE, PB-1, FSV, AVR, THTR, HTTR, HTR-10 were built and demonstrated their safety. Large GA-HTR, RSA-PBMR projects are canceled and US-NGNP project is idling. Only Chinese HTR-PM demonstrator is under construction. HTGR has long history of development. For realization and market penetration, VHTR community should look at niche market such as carbon free energy supply to industry complex, electric power for small grid, carbon free hydrogen production, power source for space colony. Technology Readiness Level should be advanced to get proper investment from industry. For this, cooperation between international R and D institutions is required. Clearly divided role between universities, research institutions, and industries will reduce complication and shorten VHTR realization day.

  12. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  13. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  14. Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

    2010-03-01

    The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the probabilistic safety assessment (PSA) domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

  15. The ENEN-III project: Technical Training on the Concepts and Design of GEN IV nuclear reactors

    International Nuclear Information System (INIS)

    Berkvens, T.; Renault, C.; Alonso, M.; Salomaa, R.; Schönfelder, C.

    2013-01-01

    Some conclusions: • Not enough training courses to cover the LO’s: – Especially GEN IV; – Many introductory courses, little specific courses; – Reach out to other partners for more courses. • Skills and Attitudes: – Much more difficult to train/measure; – To be treated in a separate project. • Use of Learning Outcomes must be promoted; • Involvement of human resources necessary for the successful implementation of the schemes: – End of project workshop

  16. Current status of NPP generation IV

    International Nuclear Information System (INIS)

    Yohanes Dwi Anggoro; Dharu Dewi; Nurlaila; Arief Tris Yuliyanto

    2013-01-01

    Today development of nuclear technology has reached the stage of research and development of Generation IV nuclear power plants (advanced reactor systems) which is an innovative development from the previous generation of nuclear power plants. There are six types of power generation IV reactors, namely: Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and Super Critical Water-cooled Reactor (SCWR). The purpose of this study is to know the development of Generation IV nuclear power plants that have been done by the thirteen countries that are members of the Gen IV International Forum (GIF). The method used is review study and refers to various studies related to the current status of research and development of generation IV nuclear power. The result of this study showed that the systems and technology on Generation IV nuclear power plants offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance and physical protection. In addition, based on the research and development experience is estimated that: SFR can be used optimally in 2015, VHTR in 2020, while NPP types GFR, LFR, MSR, and SCWR in 2025. Utilization of NPP generation IV said to be optimal if fulfill the goal of NPP generation IV, such as: capable to generate energy sustainability and promote long-term availability of nuclear fuel, minimize nuclear waste and reduce the long term stewardship burden, has an advantage in the field of safety and reliability compared to the previous generation of NPP and VHTR technology have a good prospects in Indonesia. (author)

  17. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR and GEN IV

    International Nuclear Information System (INIS)

    O'Donnell, William J.; Griffin, Donald S.

    2007-01-01

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  18. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

    Energy Technology Data Exchange (ETDEWEB)

    William J. O’Donnell; Donald S. Griffin

    2007-05-07

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  19. Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Kwon, Young Min; Kim, Tae Woon; Park, Soo Yong; Suk, Soo Dong; Lee, Kwi Lim; Lee, Yong Bum; Chang, Won Pyo; Ha, Kwi Seok; Hahn, Sang Hoon

    2010-07-01

    Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

  20. The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Bordier, Gilles; Warin, Dominique; Masson, Michel [CEA/Marcoule Direction, BP 17171 - 30207 - Bagnols-sur-Ceze Cedex (France)

    2008-07-01

    The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX{sup TM} process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO{sub 2} powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

  1. The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

    International Nuclear Information System (INIS)

    Bordier, Gilles; Warin, Dominique; Masson, Michel

    2008-01-01

    The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX TM process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO 2 powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

  2. The safety R and D for GEN-IV reactors in the European nuclear energy technology platform strategic research agenda

    International Nuclear Information System (INIS)

    Bruna, G.

    2009-01-01

    In the fall 2007 EC launched the Sustainable Nuclear Energy Technology Platform (SNE-TP). The SNE-TP governing board set-up three working groups (WG): 1) Strategic Research Agenda (SRA) WG, in charge of drafting road-maps to support research, development and demonstration for current and future NPPs; 2) Deployment Strategy (DS) WG, in charge of defining the research road-map implementation and 3) Education, Training and Knowledge management (ETKM) WG, which was aimed at issuing proposal to reinforce European education and attract young in the nuclear field. The SRA WG was mandated to prepare the SRA vision document based on the preliminary road-map sketched in the document published by the Commission earlier in 2007. The SRA WG was originally organized in 5 sub-groups covering specific topics (1) GEN II and III, III+, including Advanced LWR, 2) Advanced Fuel Cycle for waste minimization and resource optimization; 3) GEN IV Fast Systems (SFR, LFR, GFR, ADS); 4) GEN IV (V) HTR and non-electricity-production applications; 5) New Nuclear Large Research Infrastructures) and 5 other sub-groups dealing with more generic cross-cutting research activities applicable to many specific topics, namely: 1) Structural material research; 2) modeling, simulation and methods, including physical data and tools and means for qualification and validation; 3) Reactor Safety, including severe accidents and human factor; 4) Advanced Driver and Minor Actinide Fuels: science and properties; 5) Pre-normative Research, Codes and Standards.The present paper is mainly aimed at summarizing the content of the SRA Safety sub-chapter focusing on GEN-IV aspects

  3. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    International Nuclear Information System (INIS)

    Lee, Jae Han; Koo, Gyeong Hoi

    2013-01-01

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm

  4. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Koo, Gyeong Hoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm.

  5. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  6. A Qualitative Assessment of Diversion Scenarios for a GEN IV Example Sodium Fast Reactor Using the GEN IV PR and PP Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Zentner, M.D.; Coles, G.A. [PNNL, P.O. Box 999, 902 Battelle Boulvard, Richland, WA 99336 (United States); Therios, I.U. [Argonne National Lab. - ANL (United States)

    2009-06-15

    An experts working group was created in 2002 by The Generation IV International Forum for the purpose of developing an internationally accepted methodology for assessing the proliferation resistance of a nuclear energy system (NES) and its individual elements. A two year case study was performed by the working group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information to designers at various levels of details, including pre-conceptual design stage. The study analyzes the response of the ESFR entire nuclear energy system to different proliferation and theft strategies. The challenges considered comprise concealed diversion, concealed misuse and abrogation strategies. This paper describes the work done in performing a qualitative assessment of potential concealed diversion scenarios from the ESFR, and includes an evaluation of the potential effect of changes in the conversion ratio on diversion strategies. (authors)

  7. Progress reports for Gen IV sodium fast reactor activities FY 2007

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Tentner, A. M.

    2007-01-01

    for prevention of progression into severe accident conditions (prevention of core melting) or for mitigation of severe accident consequences (mitigation of the impact of core melting to protect public health and safety). Because design measures for severe accident prevention and mitigation are beyond the normal design basis, established regulatory guidelines and codes do not provide explicit identification of the design performance requirements for severe accident accommodation. The treatment of severe accidents is one of the key issues of R and D plans for the Gen IV systems in general, and for the Sodium Fast Reactor (SFR) in particular. Despite the lack of an unambiguous definition of safety approach applicable for severe accidents, there is an emerging consensus on the need for their consideration for the design. The US SFR program and Argonne National Laboratory (ANL) in particular have actively studied the potential scenarios and consequences of Hypothetical Core Disruptive Accidents (HCDA) for SFRs with oxide fuel during the Fast Flux Test Facility (FFTF) and Clinch River Breeder Reactor Plant (CRBRP) programs in the 70s and 80s. Later, the focus of the US SFR safety R and D activities shifted to the prevention of all HCDAs through passive safety features of the SFRs with metal fuel in the Integral Fast Reactor (IFR) program, and the study of severe accident consequences was de-emphasized. The goal of this paper is to provide an overview of the current SFR safety approach and the role of severe accidents in Japan and France, in preparation for an expected and more active collaboration in this area between the US, Japan, and France

  8. Generation IV nuclear reactors: Current status and future prospects

    International Nuclear Information System (INIS)

    Locatelli, Giorgio; Mancini, Mauro; Todeschini, Nicola

    2013-01-01

    Generation IV nuclear power plants (GEN IV NPPs) are supposed to become, in many countries, an important source of base load power in the middle–long term (2030–2050). Nowadays there are many designs of these NPPs but for political, strategic and economic reasons only few of them will be deployed. International literature proposes many papers and reports dealing with GEN IV NPPs, but there is an evident difference in the types and structures of the information and a general unbiased overview is missing. This paper fills the gap, presenting the state-of-the-art for GEN IV NPPs technologies (VHTR, SFR, SCWR, GFR, LFR and MSR) providing a comprehensive literature review of the different designs, discussing the major R and D challenges and comparing them with other advanced technologies available for the middle- and long-term energy market. The result of this research shows that the possible applications for GEN IV technologies are wider than current NPPs. The economics of some GEN IV NPPs is similar to actual NPPs but the “carbon cost” for fossil-fired power plants would increase the relative valuation. However, GEN IV NPPs still require substantial R and D effort, preventing short-term commercial adoption. - Highlights: • Generation IV reactors are the middle–long term technology for nuclear energy. • This paper provides an overview and a taxonomy for the designs under consideration. • R and D efforts are in the material, heat exchangers, power conversion unit and fuel. • The life cycle costs are competitive with other innovative technologies. • The hydrogen economy will foster the development of Generation IV reactors

  9. Current status of VHTR development in Japan

    International Nuclear Information System (INIS)

    Aochi, A.; Kondo, T.

    1982-01-01

    The status of the program at the beginning of fiscal 1982 is reviewed. Special emphasis is placed on the altering of the output helium temperature of the experimental VHTR to 950 0 . The modification is aimed at establishing the technical basis for post-experimental VHTR output helium temperature of 1000 0 C. Notes are given on the design of the VHTR as well as various research and development efforts in Japan on multi-purpose nuclear heat applications and HTGR technology

  10. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    International Nuclear Information System (INIS)

    Fernandez, A.; McGinley, J.; Somers, J.

    2008-01-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  11. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.; McGinley, J.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O.Box 2340, Karlsruhe, D-76125 (Germany)

    2008-07-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  12. Bolivian Rhinotragini IV: Paraeclipta gen. nov. (Coleoptera, Cerambycidae, new species and new combinations

    Directory of Open Access Journals (Sweden)

    Robin O. S. Clarke

    2011-01-01

    Full Text Available Paraeclipta gen. nov. is described to allocate five new species, and ten transferred from Eclipta Bates, 1873: P. cabrujai sp. nov.; P. clementecruzi sp. nov.; P. melgarae sp. nov.; P. tomhacketti sp. nov.; P. moscosoi sp. nov.; P. bicoloripes (Zajciw, 1965, comb. nov.; P. croceicornis (Gounelle, 1911, comb. nov.; P. flavipes (Melzer, 1922, comb. nov.; P. jejuna (Gounelle, 1911, comb. nov.; P. kawensis (Peñaherrera-Leiva & Tavakilian, 2004, comb. nov.; P. longipennis (Fisher, 1947, comb. nov.; P. rectipennis (Zajciw, 1965, comb. nov.; P. soumourouensis (Tavakilian & Peñaherrera-Leiva, 2003, comb. nov.; P. tenuis (Burmeister, 1865, comb. nov.; and P. unicoloripes (Zajciw, 1965, comb. nov. The Bolivian species are illustrated. A key to their identification and host flower records are provided.

  13. Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; De Izarra, G. [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Elter, Zs.; Pazsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goteborg, (Sweden); Verma, V.; Hellesen, C.; Jacobsson, S. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala, (Sweden); Hamrita, H.; Bakkali, M. [CEA, DRT, LIST, Sensors and Electronic Architecture Laboratory, Saclay, F-91191 Gif Sur Yvette, (France); Chapoutier, N.; Scholer, A-C.; Verrier, D. [AREVA NP, 10 rue Juliette Recamier F-69456 Lyon, (France); Cantonnet, B.; Nappe, J-C. [PHONIS France S.A.S, Nuclear Instrumentation, Avenue Roger Roncier, B.P. 520, F-19106 Brive Cedex, (France); Molinie, P.; Dessante, P.; Hanna, R.; Kirkpatrick, M.; Odic, E. [Supelec, Department of Power and Energy System, F-91192 Gif Sur Yvette, (France); Jadot, F. [CEA, DEN, DER, ASTRID Project Group, Cadarache, F-13108 Saint-Paul-lez-Durance, (France)

    2015-07-01

    The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

  14. Innovative materials for GEN IV systems and transmutation facilities (cross-cutting research project GETMAT)

    International Nuclear Information System (INIS)

    Fazio, Concetta; Rieth, Michael; Gomez Briceno, Dolores; Gessi, Alessandro; Henry, Jean; Malerba, Lorenzo

    2010-01-01

    The objectives of the 'Generation IV and Transmutation Materials' (GETMAT) project is to contribute to the development, qualification and ranking of different types of ODS steels and to qualify Ferritic/Martensitic steels in a wide irradiation condition range. The experimental approach is complemented by the development of physical models with the aim to understand and improve the predictability of the materials performance. The GETMAT consortium is composed of fourteen Research centres, nine Universities and one Utility, from eleven European countries. The R and D tasks address (i) the materials availability, fabricability, weldability and their fundamental mechanical properties, (ii) their compatibility with aggressive coolants and development of corrosion protection methods; (iii) their performance under neutron irradiation, and (iv) starting from model alloys relevant for the two classes of alloys, the development and validation of physical models. The exploitation of results to potential end-users will occur through the 'GETMAT User Group', where exchange of information with the nuclear and steel industries, international (outside Europe) Research Organisations and engineers involved in the design of the new reactors, will occur. The exploitation of results to potential end-users will occur through the G ETMAT User Group , where exchange of information with the nuclear and steel industries, international (outside Europe) Research Organisations and engineers involved in the design of the new reactors, will occur

  15. Safety design approach for JSFR toward the realization of GEN IV SFR

    International Nuclear Information System (INIS)

    Kubo, S.; Yamano, H.; Chikazawa, Y.; Shimakawa, Y.

    2013-01-01

    Conclusion: Safety Design Approach for JSFR: • Based on the safety design criteria for Generation-IV SFR • DECs, Situations practically eliminated and related design measures are identified and selected with due consideration of the safety features of SFR and the lessons learned from the TEPCO’s Fukushima Dai-ichi nuclear power plants accident Safety Design Concept of JSFR: • For failure to shutdown: Passive shutdown capability, Mitigation of core damage (Prevention of severe mechanical energy release, In-Vessel Retention) • For failure to remove heat: Prevention of significant core damage (Natural circulation DHR, Alternative cooling measures) • Containment: Prevention of sever dynamic loads by design measures (IVR, double boundary concept, inertization)

  16. Development on experimental VHTR instrumentation

    International Nuclear Information System (INIS)

    Wakayama, N.; Ara, K.; Terada, H.; Yamagishi, H.; Tomoda, T.

    1982-06-01

    This paper describes developmental works on the instrumentation of the Experimental VHTR. In the area of the nuclear instrumentation for the reactor control, high temperature fission counter-chambers have been developed. These withstood the accelerated irradiation life tests at 600 deg. C, the long term in-reactor operating test at 600 deg. C and the 800 deg. C-operating tests for several hundred hours in a simulated accident condition. Platinum-Molybdenum alloy thermocouples have been studied as a neutron-irradiation-resistant high-temperature thermocouple for the in-core temperature distribution monitoring of the VHTR in the temperature range between 1000 deg. C and 1350 deg. C. The instability problems of the Pt-5% Mo/Pt-0.1% Mo thermocouple seem to be overcome by introducing a double sheath structure and adopting a better material to the inner sheath. A local failure and abnormality monitoring method for the HTR fuel is also studied using a gas-sweeping irradiation rig for the CPF compacts. This study aims mainly at the development of a method to compensate for the dependency of the FP-release rate on the fuel temperature, the neutron flux density, the burn-up and others, in order to increase the detection sensitivity of fuel failures. (author)

  17. Hydrogen production using the sulfur-iodine cycle coupled to a VHTR: An overview

    International Nuclear Information System (INIS)

    Vitart, X.; Le Duigou, A.; Carles, P.

    2006-01-01

    The sulfur-iodine thermo-chemical cycle is considered to be one of the most promising routes for massive hydrogen production, using high temperature heat from a Generation IV VHTR. We propose here a brief overview of the main questions raised by this cycle, along with the general lines of French CEA's program

  18. A Stochastic Proof of the Resonant Scattering Kernel and its Applications for Gen IV Reactors Type

    International Nuclear Information System (INIS)

    Becker, B.; Dagan, R.; Broeders, C.H.M.; Lohnert, G.

    2008-01-01

    Monte Carlo codes such as MCNP are widely accepted as almost-reference for reactor analysis. The Monte Carlo Code should therefore use as few as possible approximations in order to produce 'experimental-level' calculations. In this study we deal with one of the most problematic approximations done in MCNP in which the resonances are ignored for the secondary neutron energy distribution, namely the change of the energy and angular direction of the neutron after interaction with a heavy isotope with pronounced resonances. The endeavour of exploiting the influence of the resonances on the scattering kernel goes back to 1944 where E. Wigner and J. Wilkins developed the first temperature dependent scattering kernel. However only in 1998, the full analytical solution for the double differential resonant dependent scattering kernel was suggested by W. Rothenstein and R. Dagan. An independent stochastic approach is presented for the first time to confirm the above analytical kernel with a complete different methodology. Moreover, by manipulating in a subtle manner the scattering subroutine COLIDN of MCNP, it is proven that this very subroutine is, to some extent, inappropriate as well as the relevant explanation in the MCNP manual. The impact of this improved resonance dependent scattering kernel on diverse types of reactors, in particular for the Generation IV innovative core design HTR, is shown to be significant. (authors)

  19. Summary of Structural Concept Development and High Temperature Structural Integrity Evaluation Technology for a Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Joo, Young Sang; Lee, Hyeong Yeon (and others)

    2008-04-15

    The economic improvement is a hot issue as one of Gen IV nuclear plant goals. It requires many researches and development works to meet the goal by securing the same level of plant safety. One of the key research items is the increase of the plant capacity with the minimum number of components and loops. Through the successful conceptual design experience for the KALIMER-600, the structural design study for a 1200MWe large capacity of sodium-cooled fast reactor has been performed to achieve the above plant size effects. The component number and reactor structural sizing were determined based on the core and fluid system design information. Several researches were performed to reduce the construction cost of NSSS in structural point of view, for example, a simplified component arrangement, concept proposals of integrated components, a high temperature LBB application technology, and an innovative in-service inspection (ISI) tool, and a computer program development of the ASME-NH design procedure of the class 1 structure and component under high temperature over 500 .deg. C. The IHTS piping arrangement was also proposed to minimize the length through the properly locating the SG and pump by 126m. Further studies of these concepts are required to confirm on the fabricability and the structural integrity for the operating and design loads. The proposed concepts will be optimized to a unified conceptual design through several trade-off studies.

  20. Status of the French R/D program on the severe accident issue to develop Gen IV SFRs - 15373

    International Nuclear Information System (INIS)

    Serre, F.; Bertrand, F.; Journeau, C.; Suteau, C.; Verwaede, D.; Schmitt, D.; Farges, B.

    2015-01-01

    The ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration) is a technological demonstrator designed by the CEA with its industrial partners, with very high levels of requirements. In the ASTRID project, the safety objectives are to prevent core melting, in particular by the development of an innovative core with complementary safety prevention devices, and to enhance the reactor resistance to severe accident by design. To mitigate the consequences of hypothetical core melting situations, specific dispositions or mitigation devices will be added to the core and to the reactor. It is also required to provide a robust safety demonstration (with high level of confidence). Therefore a new approach for severe accident issue has been defined: to the well-known 'lines of defense' method, a 'lines of mitigation' method is added. To meet these ASTRID, or future SFR, requirements, a large R/D program was launched in the Severe Accident domain, with a large number of partners. This paper will present the status of the CEA R/D related to the SFR Severe Accident issue, the collaboration framework (with industrial partners and R/D foreign organizations), and the future R/D plans to support the ASTRID project and possible developments for future Gen IV commercial SFR. (authors)

  1. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  2. Enhanced radiation resistance through interface modification of nano-structured steels for Gen IV in-core applications

    International Nuclear Information System (INIS)

    Jang, Jinsung; Kang, Suk Hoon; Kim, Min Chul

    2013-06-01

    This project is to increase radiation tolerance of candidate alloys for Gen IV core component through the optimization of grain size and grain boundary characteristics. The focus is on nanocrystalline metal alloys with a fcc crystal structure. The long-term goal is to design and develop bulk nanostructured austenitic steels with enhanced void swelling resistance and substantial ductility, and to enhance their creep resistance at elevated temperatures via grain boundary engineering. An austenitic stainless steel, HT-UPS (high temperature ultra-fine precipitates strengthened) was developed at ORNL, and is expected to show enhanced void swelling resistance through the trapping of point defects at nanometer-sized carbides. Reducing the grain size and increasing the fraction-induced point defects (due to the increased sink area of the grain boundaries), to make bubble nucleation at the boundaries less likely (by reducing the fraction of high-energy boundaries), and to improve the strength and ductility under radiation by producing a higher density of nanometer sized carbides on the boundaries

  3. Design and Selection of Innovative Primary Circulation Pumps for GEN-IV Lead Fast Reactors

    Directory of Open Access Journals (Sweden)

    Walter Borreani

    2017-12-01

    Full Text Available Although Lead-cooled Fast Reactor (LFR is not a new concept, it continues to be an example of innovation in the nuclear field. Recently, there has been strong interest in liquid lead (Pb or liquid lead–bismuth eutectic (LBE both critical and subcritical systems in a relevant number of Countries, including studies performed in the frame of GENERATION-IV initiative. In this paper, the theoretical and computational findings for three different designs of Primary Circulation Pump (PCP evolving liquid lead (namely the jet pump, the Archimedean pump and the blade pump are presented with reference to the ALFRED (Advanced Lead Fast Reactor European Demonstrator design. The pumps are first analyzed from the theoretical point of view and then modeled with a 3D CFD code. Required design performance of the pumps are approximatively around an effective head of 2 bar with a mass flow rate of 5000 kg/s. Taking into account the geometrical constraints of the reactor and the fluid dynamics characteristics of the molten lead, the maximum design velocity for molten lead fluid flow of 2 m/s may be exceeded giving rise to unacceptable erosion phenomena of the blade or rotating component of the primary pumping system. For this reason a deep investigation of non-conventional axial pumps has been performed. The results presented shows that the design of the jet pump looks like beyond the current technological feasibility while, once the mechanical challenges of the Archimedean (screw pump and the fluid-dynamic issues of the blade pump will be addressed, both could represent viable solutions as PCP for ALFRED. Particularly, the blade pump shows the best performance in terms of pressure head generated in normal operation conditions as well as pressure drop in locked rotor conditions. Further optimizations (mainly for what the geometrical configuration is concerned are still necessary.

  4. Status of the Gen-IV Proliferation Resistance and Physical Protection (PRPP) Evaluation Methodology

    International Nuclear Information System (INIS)

    Whitlock, J.; Bari, R.; Peterson, P.; Padoani, F.; Cojazzi, G.G.M.; Renda, G.; ); Cazalet, J.; Haas, E.; Hori, K.; Kawakubo, Y.; Chang, S.; Kim, H.; Kwon, E.-H.; Yoo, H.; Chebeskov, A.; Pshakin, G.; Pilat, J.F.; Therios, I.; Bertel, E.

    2015-01-01

    Methodologies have been developed within the Generation IV International Forum (GIF) to support the assessment and improvement of system performance in the areas safeguards, security, economics and safety. Of these four areas, safeguards and security are the subjects of the GIF working group on Proliferation Resistance and Physical Protection (PRPP). Since the PRPP methodology (now at Revision 6) represents a mature, generic, and comprehensive evaluation approach, and is freely available on the GIF public website, several non-GIF technical groups have chosen to utilize the PRPP methodology for their own goals. Indeed, the results of the evaluations performed with the methodology are intended for three types of generic users: system designers, programme policy makers, and external stakeholders. The PRPP Working Group developed the methodology through a series of demonstration and case studies. In addition, over the past few years various national and international groups have applied the methodology to inform nuclear energy system designs, as well as to support the development of approaches to advanced safeguards. A number of international workshops have also been held which have introduced the methodology to design groups and other stakeholders. In this paper we summarize the technical progress and accomplishments of the PRPP evaluation methodology, including applications outside GIF, and we outline the PRPP methodology's relationship with the IAEA's INPRO methodology. Current challenges with the efficient implementation of the methodology are outlined, along with our path forward for increasing its accessibility to a broader stakeholder audience - including supporting the next generation of skilled professionals in the nuclear non-proliferation field. (author)

  5. FBR/VHTR deployment scenarios in Japan

    International Nuclear Information System (INIS)

    Richards, Matt; Kunitomi, Kazuhiko

    2008-01-01

    Co-deployment of Fast Breeder Reactors (FBRs) and Very High Temperature Reactors (VHTRs) can be used as the nuclear technologies to meet a significant portion of Japan's future energy demands. The FBR provides the fissile fuel for energy security and sustainability, and can be used to provide a significant portion of the electricity demand. The VHTR can provide flexible energy outputs (electricity, hydrogen, and high-temperature heat) with high efficiency, can operate with a wide variety of fuel cycles, and can be sited at locations that have limited availability of cooling water. These features, combined with its passive safety and high degree of proliferation resistance, make the VHTR an ideal complement for co-deployment with the FBR in Japan and also a very low-risk technology of export to foreign countries. In addition to hydrogen production, the high-temperature thermal energy produced by the VHTR fleet can be used for a wide variety of process-heat applications, and the VHTR can play a key role for significantly reducing greenhouse-gas emissions. This paper describes assessments for deploying FBRs and VHTRs in Japan using a closed fuel cycle, with the FBRs supplying the fissile material to sustain the combined FBR/VHTR fleet. (author)

  6. Development of Essential Technology for VHTR

    International Nuclear Information System (INIS)

    Kim, Yong Wan; Koo, G. H.; Kim, D. H.

    2009-04-01

    The research tasks performed in this project can be classified into five categories; high temperature material of VHTR reactor and components for hydrogen production, the nuclear graphite for the core material, the essential technologies for VHTR components, Process Heat Exchanger (PHE) fabrication, and gas loop for PHE verification tests. Research tasks on high temperature materials of VHTR reactor and components include creep properties of super alloy for high temperature components, properties of a modified 9Cr-1Mo alloy, fabrication and properties of in-core ceramic composites, and corrosion properties of the materials for the sulfuric acid decomposer. The technologies of graphitization evaluation, nondestructive defect detection, and impurity analysis were developed in field of nuclear graphites. The properties of graphites were evaluated by tests using small specimen test. The abroad status of graphite machining was reviewed. Review about the status of VHTR components, structural sizing and analysis for hot gas duct, thermal sizing of IHX were performed in the field of the essential technologies for VHTR components. The surface modification process with ion beam mixing was optimized and evaluated for the fabrication of process heat exchanger (PHE). The secondary sulfuric acid loop was designed and constructed in the gas loop. The lab-scale PHE test was performed in the gas loop. In addition, the conceptual design of the mid-size helium loop was performed in the next stage of this project

  7. Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

    International Nuclear Information System (INIS)

    Stauff, N.

    2011-01-01

    Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

  8. Salt Selection for the LS-VHTR

    International Nuclear Information System (INIS)

    Williams, D.F.; Clarno, K.T.

    2006-01-01

    Molten fluorides were initially developed for use in the nuclear industry as the high temperature fluid-fuel for a Molten Salt Reactor (MSR). The Office of Nuclear Energy is exploring the use of molten fluorides as a primary coolant (rather than helium) in an Advanced High Temperature Reactor (AHTR) design, also know as the Liquid-Salt cooled Very High Temperature Reactor (LS-VHTR). This paper provides a review of relevant properties for use in evaluation and ranking of candidate coolants for the LS-VHTR. Nuclear, physical, and chemical properties were reviewed and metrics for evaluation are recommended. Chemical properties of the salt were examined for the purpose of identifying factors that effect materials compatibility (i.e., corrosion). Some preliminary consideration of economic factors for the candidate salts is also presented. (authors)

  9. Preliminary shielding analysis of VHTR reactors

    International Nuclear Information System (INIS)

    Flaspoehler, Timothy M.; Petrovic, Bojan

    2011-01-01

    Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)

  10. Gens Fera. The Wild men in the system of border decoration of the Bible of Wenceslas IV

    Czech Academy of Sciences Publication Activity Database

    Studničková, Milada

    2014-01-01

    Roč. 62, č. 3 (2014), s. 214-239 ISSN 0049-5123 R&D Projects: GA ČR GA13-39192S Institutional support: RVO:68378033 Keywords : book illumination * Bible of Wenceslas IV Subject RIV: AL - Art, Architecture, Cultural Heritage http://www.umeni-art.cz/admin/fileGet.aspx?v=issue-issue-2267-category-2268-paragraph-2269-pdf&l=cz

  11. A comparison of the risk measures between VHTR and LWR

    International Nuclear Information System (INIS)

    Han, Seok-Jung; Yang, Joon-Eon; Lee, Won-Jea

    2007-01-01

    Because the safety characteristics of a very high temperature reactor (VHTR) are different to that of light water reactors (LWRs), it is necessary to develop an adequate probabilistic safety assessment (PSA) methodology in order to perform a risk assessment. The inherent safety features of the VHTR are (1) simplified safety functions (2) the absence of the large release of radioactive materials such as a severe accident in LWRs. The PSA methodology for LWRs cannot be directly applied in a VHTR PSA. This paper proposes a PSA methodology for a VHTR. The essential point of the proposed methodology is to define end states of accident sequences in order to establish the risk measures for a VHTR PSA. This paper compares them with that for LWRs to discuss the differences of them

  12. Development of a Methodology for VHTR Accident Consequence Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    The substitution of the VHTR for burning fossil fuels conserves these hydrocarbon resources for other uses and eliminates the emissions of greenhouse. In Korea, for these reasons, constructing the VHTR plan for hydrogen production is in progress. In this study, the consequence analysis for the off-site releases of radioactive materials during severe accidents has been performed using the level 3 PRA technology. The offsite consequence analysis for a VHTR using the MACCS code has been performed. Since the passive system such as the RCCS(Reactor Cavity Cooling System) are equipped, the frequency of occurrence of accidents has been evaluated to be very low. For further study, the assessment for characteristic of VHTR safety system and precise quantification of its accident scenarios is expected to conduct more certain consequence analysis. This methodology shown in this study might contribute to enhancing the safety of VHTR design by utilizing the results having far lower effect on the environment than the LWRs.

  13. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  14. Experimental Measurement of Flow Phenomena in a VHTR Lower Plenum Model

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Keith G. Condie; Glenn E. McCreery; Donald M. McEligot; Robert J. Pink

    2006-06-01

    The Very-High-Temperature Reactor (VHTR) is one of six reactor technologies chosen for further development by the Generation IV International Forum. In addition this system is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. In preparation for the thermal-hydraulics and safety analyses that will be required to confirm the performance of the NGNP, work has begun on readying the computational tools that will be needed to predict the thermal-hydraulics conditions and safety margins of the reactor design. Meaningful feasibility studies for VHTR designs will require accurate, reliable predictions of material temperatures which depend upon the thermal convection in the coolant channels of the core and other components. Unfortunately, one-dimensional system codes for gas-cooled reactors typically underpredict these temperatures, particularly for reduced power operations and hypothesized accident scenarios. Likewise, most turbulence models in general-purpose CFD codes also underpredict these temperatures. Matched-Index-of-Refraction (MIR) fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical VHTR plenum geometries in the limiting case of negligible buoyancy and constant fluid properties.

  15. A Qualitative Assessment Of Diversion Scenarios For A Example Sodium Fast Reactor Using The Gen IV PR And PP Methodology

    International Nuclear Information System (INIS)

    Zentner, Michael D.

    2008-01-01

    A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

  16. A Qualitative Assessment of Diversion Scenarios for an Example Sodium Fast Reactor Using the GEN IV PR and PP Methodology

    International Nuclear Information System (INIS)

    Zentner, Michael D.; Coles, Garill A.; Therios, Ike

    2012-01-01

    FAST REACTORS;NUCLEAR ENERGY;NUCLEAR MATERIALS MANAGEMENT;PROLIFERATION;SAFEGUARDS;THEFT; A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

  17. A preliminary neutronic evaluation and depletion study of VHTR and LS-VHTR reactors using the codes: WIMSD5 and MCNPX

    International Nuclear Information System (INIS)

    Silva, Fabiano C.; Pereira, Claubia; Costa, Antonella Lombardi; Veloso, Maria Auxiliadora Fortini

    2009-01-01

    It is expected that, in the future, besides electricity generation, reactors should also develop secondary activities, such as hydrogen generation and seawater desalinization. Generation IV reactors are expected to possess special characteristics, like high safety, minimization of radioactive rejects amount and ability to use reprocessed fuel with non-proliferating projects in their cycles. Among the projects of IV generation reactors available nowadays, the (High Temperature Reactors) HTR, are highlighted due to these desirable characteristics. Under such circumstances, such reactor may be able to have significant higher thermal power ratings to be used for hydrogen production, without loose of safety, even in an emergency. For this work, we have chosen two HTR concepts of a prismatic reactor: (Very High Temperature Reactor) VHTR and the (Liquid Salted -Very High Temperature Reactor) LS-VHTR. The principal difference between them is the coolant. The VHTR uses helium gas as a coolant and have a burnup of 101,661 MWd/THM while the LS-VHTR uses low-pressure liquid coolant molten fluoride salt with a boiling point near 1500 de C working at 155,946 MWd/THM. The ultimate power output is limited by the capacity of the passive decay system; this capacity is limited by the reactor vessel temperature. The goal was to evaluate the neutronic behavior and fuel composition during the burnup using the codes (Winfrith Improved Multi-Group Scheme) WIMSD5 and the MCNPX2.6. The first, deterministic and the second, stochastic. For both reactors, burned fuel type 'C' coming from Angra-I nuclear plant, in Brazil, was used with 3.1% of initial enrichment, burnup to 33,000 MWd/THM using the ORIGEN2.1 code, divided in three steps of 11,000 MWd/THM, with an average density power of 37.75 MWd/THM and 5 years of cooling in pool. Finally, the fuel was reprocessed by Purex technique extracting 99.9% of Pu, and the desired amount of fissile material (15%) to achieve the final mixed oxide was

  18. Economic, energy and GHG emissions performance evaluation of a WhisperGen Mk IV Stirling engine μ-CHP unit in a domestic dwelling

    International Nuclear Information System (INIS)

    Conroy, G.; Duffy, A.; Ayompe, L.M.

    2014-01-01

    Highlights: • The performance of a Stirling engine MK IV micro-CHP unit was evaluated in a domestic dwelling in Ireland. • The performance of the micro-CHP was compare to that of a condensing gas boiler. • The micro-CHP unit resulted in an annual cost saving of €180 compared to the condensing gas boiler. • Electricity imported from the grid decreased by 20.8% while CO 2 emissions decreased by 16.1%. • The micro-CHP unit used 2889 kW h of gas more than the condensing gas boiler during one year of operation. - Abstract: This paper presents an assessment of the energy, economic and greenhouse gas emissions performances of a WhisperGen Mk IV Stirling engine μ-CHP unit for use in a conventional house in the Republic of Ireland. The energy performance data used in this study was obtained from a field trial carried out in Belfast, Northern Ireland during the period June 2004–July 2005 by Northern Ireland Electricity and Phoenix Gas working in collaboration with Whispertech UK. A comparative performance analysis between the μ-CHP unit and a condensing gas boiler revealed that the μ-CHP unit resulted in an annual cost saving of €180 with an incremental simple payback period of 13.8 years when compared to a condensing gas boiler. Electricity imported from the grid decreased by 20.8% while CO 2 emissions decreased by 16.1%. The μ-CHP unit used 2889 kW h of gas more than the condensing gas boiler

  19. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  20. The web-enabled database of JRC-EC, a useful tool for managing European Gen IV materials data

    International Nuclear Information System (INIS)

    Over, H.H.; Dietz, W.

    2008-01-01

    Materials and document databases are important tools to conserve knowledge and experimental materials data of European R and D projects. A web-enabled application guarantees a fast access to these data. In combination with analysis tools the experimental data are used for e.g. mechanical design, construction and lifetime predictions of complex components. The effective and efficient handling of large amounts of generic and detailed materials data with regard to properties related to e.g. fabrication processes, joining techniques, irradiation or aging is one of the basic elements of data management within ongoing nuclear safety and design related European research projects and networks. The paper describes the structure and functionality of Mat-DB and gives examples how these tools can be used for the management and evaluation of materials data of European (national or multi-national) R and D activities or future reactor types such as the EURATOM FP7 Generation IV reactor types or the heavy liquid metals cooled reactor

  1. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  2. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  3. Small angle neutron scattering study of nano sized microstructure in Fe-Cr ODS steels for gen IV in-core applications.

    Science.gov (United States)

    Han, Young-Soo; Mao, Xiadong; Jang, Jinsung

    2013-11-01

    The nano-sized microstructures in Fe-Cr oxide dispersion strengthened steel for Gen IV in-core applications were studied using small angle neutron scattering. The oxide dispersion strengthened steel was manufactured through hot isostatic pressing with various chemical compositions and fabrication conditions. Small angle neutron scattering experiments were performed using a 40 m small angle neutron scattering instrument at HANARO. Nano sized microstructures, namely, yttrium oxides and Cr-oxides were quantitatively analyzed by small angle neutron scattering. The yttrium oxides and Cr-oxides were also observed by transmission electron microscopy. The microstructural analysis results from small angle neutron scattering were compared with those obtained by transmission electron microscopy. The effects of the chemical compositions and fabrication conditions on the microstructure were investigated in relation to the quantitative microstructural analysis results obtained by small angle neutron scattering. The volume fraction of Y-oxide increases after fabrication, and this result is considered to be due to the formation of non-stochiometric Y-Ti-oxides.

  4. Whole core transport calculation for the VHTR hexagonal core

    International Nuclear Information System (INIS)

    Cho, J. Y.; Kim, K. S.; Lee, C. C.; Joo, H. G.

    2007-01-01

    occupies several cells has no influence on the MOC transport calculation, but it introduces a near singular matrix for a CMFD formulation. This void problem is resolved by introducing a lumped CMR scheme in which the void cells are collapsed to one equation to produce one rebalancing factor. The lumped CMR scheme causes the original CMFD equation not to converge. Therefore, the CMFD calculation is stopped if the residual error of the CMFD solution does not reduce, and return to the radial MOC transport calculation.In the comparison of the computational result with the MCNP code for the VHTR 2-D core problem, DeCART shows about 200 ∼ 500 pcm difference in the eigenvalue, and less than 1.0 % difference in the assembly power distribution. For the computing time, DeCART takes less than 5 hours on a PENTIUM-IV 3.0 GHz personal computer. Those results indicated that the hexagonal module of the DeCART code worked very well within an affordable computing time

  5. Scaled Experimental Modeling of VHTR Plenum Flows

    Energy Technology Data Exchange (ETDEWEB)

    ICONE 15

    2007-04-01

    Abstract The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. Various scaled heated gas and water flow facilities were investigated for modeling VHTR upper and lower plenum flows during the decay heat portion of a pressurized conduction-cooldown scenario and for modeling thermal mixing and stratification (“thermal striping”) in the lower plenum during normal operation. It was concluded, based on phenomena scaling and instrumentation and other practical considerations, that a heated water flow scale model facility is preferable to a heated gas flow facility and to unheated facilities which use fluids with ranges of density to simulate the density effect of heating. For a heated water flow lower plenum model, both the Richardson numbers and Reynolds numbers may be approximately matched for conduction-cooldown natural circulation conditions. Thermal mixing during normal operation may be simulated but at lower, but still fully turbulent, Reynolds numbers than in the prototype. Natural circulation flows in the upper plenum may also be simulated in a separate heated water flow facility that uses the same plumbing as the lower plenum model. However, Reynolds number scaling distortions will occur at matching Richardson numbers due primarily to the necessity of using a reduced number of channels connected to the plenum than in the prototype (which has approximately 11,000 core channels connected to the upper plenum) in an otherwise geometrically scaled model. Experiments conducted in either or both facilities will meet the objectives of providing benchmark data for the validation of codes proposed for NGNP designs and safety studies, as well as providing a better understanding of the complex flow phenomena in the plenums.

  6. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  7. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Behafarid, F.; Shaver, D. R. [Rensselaer Polytechnic Inst., Troy, NY (United States); Bolotnov, I. A. [North Carolina State Univ., Raleigh, NC (United States); Jansen, K. E. [Univ. of Colorado, Boulder, CO (United States); Antal, S. P.; Podowski, M. Z. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  8. Application of Network Analysis Method to VHTR core

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl

    2012-01-01

    A Very High Temperature Reactor (VHTR) is currently envisioned as a promising future reactor concept because of its high-efficiency and capability of generating hydrogen. Prismatic Modular Reactor (PMR) is one of the main VHTR concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However their shape could be changed by neutron damage during the reactor operation and the shape change can makes the gaps between the blocks inducing bypass flow. Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Therefore, fast, flexible and reliable code is required to predict the flow distribution corresponding to the various bypass gap distribution. Consequently in this study, the flow network analysis method is applied to analyze the core flow of VHTR. The applied method was validated by comparing with SNU VHTR multiblock experiment. As a result, the calculated results show good agreements with experimental data although computational time and cost of the developed code was very small

  9. Evaluation of the DRAGON code for VHTR design analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  10. Evaluation of the DRAGON code for VHTR design analysis

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-01

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR

  11. Status of experimental data for the VHTR core design

    Energy Technology Data Exchange (ETDEWEB)

    Park, Won Seok; Chang, Jong Hwa; Park, Chang Kue

    2004-05-01

    The VHTR (Very High Temperature Reactor) is being emerged as a next generation nuclear reactor to demonstrate emission-free nuclear-assisted electricity and hydrogen production. The VHTR could be either a prismatic or pebble type helium cooled, graphite moderated reactor. The final decision will be made after the completion of the pre-conceptual design for each type. For the pre-conceptual design for both types, computational tools are being developed. Experimental data are required to validate the tools to be developed. Many experiments on the HTGR (High Temperature Gas-cooled Reactor) cores have been performed to confirm the design data and to validate the design tools. The applicability and availability of the existing experimental data have been investigated for the VHTR core design in this report.

  12. VHTR engineering design study: intermediate heat exchanger program. Final report

    International Nuclear Information System (INIS)

    1976-11-01

    The work reported is the result of a follow-on program to earlier Very High Temperature Reactor (VHTR) studies. The primary use of the VHTR is to provide heat for various industrial processes, such as hydrocarbon reforming and coal gasification. For many processes the use of an intermediate heat transfer barrier between the reactor coolant and the process is desirable; for some processes it is mandatory. Various intermediate heat exchanger (IHX) concepts for the VHTR were investigated with respect to safety, cost, and engineering design considerations. The reference processes chosen were steam-hydrocarbon reforming, with emphasis on the chemical heat pipe, and steam gasification of coal. The study investigates the critically important area of heat transfer between the reactor coolant, helium, and the various chemical processes

  13. Computational Fluid Dynamic Analysis of the VHTR Lower Plenum Standard Problem

    International Nuclear Information System (INIS)

    Johnson, Richard W.; Schultz, Richard R.

    2009-01-01

    The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 C to perhaps 1000 C. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U.S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present report presents results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made

  14. Corrosion of structural materials for Generation IV systems

    International Nuclear Information System (INIS)

    Balbaud-Celerier, F.; Cabet, C.; Courouau, J.L.; Martinelli, L.; Arnoux, P.

    2009-01-01

    The Generation IV International Forum aims at developing future generation nuclear energy systems. Six systems have been selected for further consideration: sodium-cooled fast reactor (SFR), gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR) and very high temperature reactor (VHTR). CEA, in the frame of a national program, of EC projects and of the GIF, contributes to the structural materials developments and research programs. Particularly, corrosion studies are being performed in the complex environments of the GEN IV systems. As a matter of fact, structural materials encounter very severe conditions regarding corrosion concerns: high temperatures and possibly aggressive chemical environments. Therefore, the multiple environments considered require also a large diversity of materials. On the other hand, the similar levels of working temperatures as well as neutron spectrum imply also similar families of materials for the various systems. In this paper, status of the research performed in CEA on the corrosion behavior of the structural material in the different environments is presented. The materials studied are either metallic materials as austenitic (or Y, La, Ce doped) and ferrito-martensitic steels, Ni base alloys, ODS steels, or ceramics and composites. In all the environments studied, the scientific approach is identical, the objective being in all cases the understanding of the corrosion processes to establish recommendations on the chemistry control of the coolant and to predict the long term behavior of the materials by the development of corrosion models. (author)

  15. VHTR Construction Ripple Effect using Inter-Industry Analysis

    International Nuclear Information System (INIS)

    Lee, T. H.; Lee, K. Y.; Shin, Y. J.

    2015-01-01

    As a part of a VHTR economic analysis, we have studied the VHTR construction cost and operation and maintenance cost. However, it is somewhat difficult to expect the exact cost due to insufficient reference data and experience. As a result, we propose quantitative analysis techniques for ripple effects such as the production inducement effect, added value inducement effect, and employment inducement effect for VHTR 600MWt x 4 module construction and operation ripple effect based on NOAK. This paper presents a new method for the ripple effect and preliminary ripple effect consequence. We proposed a ripple effect analysis method using a time series and inter-industry table. As a result, we can predict that a 600MWth x 4 module VHTR reactor construction will bring about a 43771 employment effect, 24160 billion KRW production effect, and 4472 billion added value effect for 22 years. It is necessary to use the sub-account values of an inter-industry table to obtain a more precise effect result. However, the methodology can be applied with minor modification to another reactor type.

  16. VHTR Construction Ripple Effect using Inter-Industry Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, T. H.; Lee, K. Y.; Shin, Y. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a part of a VHTR economic analysis, we have studied the VHTR construction cost and operation and maintenance cost. However, it is somewhat difficult to expect the exact cost due to insufficient reference data and experience. As a result, we propose quantitative analysis techniques for ripple effects such as the production inducement effect, added value inducement effect, and employment inducement effect for VHTR 600MWt x 4 module construction and operation ripple effect based on NOAK. This paper presents a new method for the ripple effect and preliminary ripple effect consequence. We proposed a ripple effect analysis method using a time series and inter-industry table. As a result, we can predict that a 600MWth x 4 module VHTR reactor construction will bring about a 43771 employment effect, 24160 billion KRW production effect, and 4472 billion added value effect for 22 years. It is necessary to use the sub-account values of an inter-industry table to obtain a more precise effect result. However, the methodology can be applied with minor modification to another reactor type.

  17. Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-392

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Izarra, G. de [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Elter, Zs. [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goeteborg (Sweden); Verma, V. [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Uppsala University, Division of Applied Nuclear Physics, Box 516, SE-75120 Uppsala (Sweden); Hamrita, H.; Bakkali, M. [CEA, DRT, LIST, Metrology, Instrumentation and Information Department, Saclay, 91191 Gif-sur-Yvette (France); Chapoutier, N.; Scholer, A.C.; Verrier, D. [AREVA NP, 10 rue Juliette Recamier F-69456 Lyon (France); Hellesen, C.; Jacobsson, S. [Uppsala University, Division of Applied Nuclear Physics, Box 516, SE-75120 Uppsala (Sweden); Pazsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goeteborg (Sweden); Cantonnet, B.; Nappe, J.C. [PHOTONIS France, Nuclear Instrumentation, 19100 Brive-la-Gaillarde (France); Molinie, P.; Dessante, P.; Hanna, R.; Kirkpatrick, M.; Odic, E. [Supelec, Energy Department, 3 rue Joliot-Curie, 91191 Gif-sur-Yvette (France)

    2015-07-01

    France has a long experience of about 50 years in designing, building and operating sodium-cooled fast reactors (SFR) such as RAPSODIE, PHENIX and SUPER PHENIX. Fast reactors feature the double capability of reducing nuclear waste and saving nuclear energy resources by burning actinides. Since this reactor type is one of those selected by the Generation IV International Forum, the French government asked, in the year 2006, CEA, namely the French Alternative Energies and Atomic Energy Commission, to lead the development of an innovative GEN-IV nuclear- fission power demonstrator. The major objective is to improve the safety and availability of an SFR. The neutron flux monitoring (NFM) system of any reactor must, in any situation, permit both reactivity control and power level monitoring from startup to full power. It also has to monitor possible changes in neutron flux distribution within the core region in order to prevent any local melting accident. The neutron detectors will have to be installed inside the reactor vessel because locations outside the vessel will suffer from severe disadvantages; radially the neutron shield that is also contained in the reactor vessel will cause unacceptable losses in neutron flux; below the core the presence of a core-catcher prevents from inserting neutron guides; and above the core the distance is too large to obtain decent neutron signals outside the vessel. Another important point is to limit the number of detectors placed in the vessel in order to alleviate their installation into the vessel. In this paper, we show that the architecture of the NFM system will rely on high-temperature fission chambers (HTFC) featuring wide-range flux monitoring capability. The definition of such a system is presented and the justifications of technological options are brought with the use of simulation and experimental results. Firstly, neutron-transport calculations allow us to propose two in-vessel regions, namely the above-core and under

  18. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Limaiem, I.

    2006-12-01

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  19. ANTARES: The HTR/VHTR project at Framatome ANP

    International Nuclear Information System (INIS)

    Gauthier, Jean-Claude; Brinkmann, Gerd; Copsey, Bernie; Lecomte, Michel

    2006-01-01

    Framatome ANP is developing a very high temperature reactor (VHTR), relying on its previous experience with high temperature reactor concepts, from its participation in the MODUL and the GT-MHR designs. While being a major actor in the nuclear reactor business with proven light water technology, AREVA wishes to be ready to meet the new challenges calling for small grid requirements, high temperature process heat and cogeneration. The Framatome ANP VHTR design for electricity production is based on an indirect cycle coupled to an 'off-the-shelf' combined cycle gas turbine. Although direct cycle HTRs are being promoted for their high efficiency, preliminary evaluations show that the Framatome ANP design efficiency is on par with a direct cycle while avoiding power generation system (PGS) developments and keeping the PGS contamination free. Moreover, the nuclear heat source of the indirect cycle could also be used to meet the heat supplies from a standard design for multiple applications

  20. Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-12-21

    The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

  1. Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

    International Nuclear Information System (INIS)

    Kawaji, Masahiro; Valentin, Francisco I.; Artoun, Narbeh; Banerjee, Sanjoy; Sohal, Manohar; Schultz, Richard; McEligot, Donald M.

    2015-01-01

    The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

  2. A reliability assessment methodology for the VHTR passive safety system

    International Nuclear Information System (INIS)

    Lee, Hyungsuk; Jae, Moosung

    2014-01-01

    The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

  3. Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life VHTR Configurations: Designs, Advantages and Limitations

    International Nuclear Information System (INIS)

    Tsvetkov, Pavel V.

    2009-01-01

    This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

  4. VHTR core modeling: coupling between neutronic and thermal-hydraulics

    International Nuclear Information System (INIS)

    Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

    2005-01-01

    Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

  5. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

    1983-03-01

    Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

  6. Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

    2005-12-01

    The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

  7. Aqueous alteration of VHTR fuels particles under simulated geological conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ait Chaou, Abdelouahed, E-mail: aitchaou@subatech.in2p3.fr; Abdelouas, Abdesselam; Karakurt, Gökhan; Grambow, Bernd

    2014-05-01

    Very High Temperature Reactor (VHTR) fuels consist of the bistructural-isotropic (BISO) or tristructural-isotropic (TRISO)-coated particles embedded in a graphite matrix. Management of the spent fuel generated during VHTR operation would most likely be through deep geological disposal. In this framework we investigated the alteration of BISO (with pyrolytic carbon) and TRISO (with SiC) particles under geological conditions simulated by temperatures of 50 and 90 °C and in the presence of synthetic groundwater. Solid state (scanning electron microscopy (SEM), micro-Raman spectroscopy, electron probe microanalyses (EPMA) and X-ray photoelectron spectroscopy (XPS)) and solution analyses (ICP-MS, ionique chromatography (IC)) showed oxidation of both pyrolytic carbon and SiC at 90 °C. Under air this led to the formation of SiO{sub 2} and a clay-like Mg–silicate, while under reducing conditions (H{sub 2}/N{sub 2} atmosphere) SiC and pyrolytic carbon were highly stable after a few months of alteration. At 50 °C, in the presence and absence of air, the alteration of the coatings was minor. In conclusion, due to their high stability in reducing conditions, HTR fuel disposal in reducing deep geological environments may constitute a viable solution for their long-term management.

  8. Gas cooled thermal reactors with high temperatures (VHTR)

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.; Vasile, A.

    2014-01-01

    VHTR is one of the 6 concepts retained for the 4. generation of nuclear reactors, it is an upgraded version of the HTR-type reactor (High Temperature Reactors). 5 HTR reactors were operated in the world in the eighties, now 2 experimental HTR are working in China and Japan and 2 HTR with an output power of 100 MWe are being built in China. The purpose of the VHTR is to provide an helium at very high temperatures around 1000 Celsius degrees that could be used directly in a thermochemical way to produce hydrogen for instance. HTR reactors are interesting in terms of safety but it does not optimise the consumption of uranium and the production of wastes. This article presents a brief historical account of HTR-type reactors and their main design and safety features. The possibility of using HTR to burn plutonium is also presented as well as the possibility of closing the fuel cycle and of using thorium-uranium fuel. (A.C.)

  9. A Quantitative Accident Sequence Analysis for a VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jintae; Lee, Joeun; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    In Korea, the basic design features of VHTR are currently discussed in the various design concepts. Probabilistic risk assessment (PRA) offers a logical and structured method to assess risks of a large and complex engineered system, such as a nuclear power plant. It will be introduced at an early stage in the design, and will be upgraded at various design and licensing stages as the design matures and the design details are defined. Risk insights to be developed from the PRA are viewed as essential to developing a design that is optimized in meeting safety objectives and in interpreting the applicability of the existing demands to the safety design approach of the VHTR. In this study, initiating events which may occur in VHTRs were selected through MLD method. The initiating events were then grouped into four categories for the accident sequence analysis. Initiating events frequency and safety systems failure rate were calculated by using reliability data obtained from the available sources and fault tree analysis. After quantification, uncertainty analysis was conducted. The SR and LR frequency are calculated respectively 7.52E- 10/RY and 7.91E-16/RY, which are relatively less than the core damage frequency of LWRs.

  10. Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

    International Nuclear Information System (INIS)

    Cai, Li

    2014-01-01

    In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

  11. Designing a TAC thermometer from a VHTR graphite structure

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James A., E-mail: James.Smith@INL.gov; Kotter, Dale, E-mail: James.Smith@INL.gov [Fuel Performance and Design, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Garrett, Steven L.; Ali, Randall A. [Graduate Program in Acoustics, Penn State University, State College, PA (United States)

    2015-03-31

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. Very High Temperature Reactors are pushing the in core temperatures even higher. A unique sensing approach will be discussed to address the necessary high temperature measurements. Thermoacoustic thermometry exploits high temperatures and uses materials that are immune to the effects of ionizing radiation to create a temperature sensor that is self-powered and wireless. In addition, the form-factor for the Thermoacoustic Thermometer (TACT) can be designed to be integrated within common in-pile structures. There are no physical moving parts required for TACT and the sensor is self-powered, as it uses the nuclear fuel for its heat source. TACT data will be presented from a laboratory prototype mimicking the design necessary for a VHTR graphite structure.

  12. Evaluation of nickel-based materials for VHTR heat exchanger

    International Nuclear Information System (INIS)

    Burlet, H.; Gentzbittel, J.M.; Cabet, C.; Lamagnere, P.; Blat, M.; Renaud, D.; Dubiez-Le Goff, S.; Pierron, D.

    2008-01-01

    Two available conventional nickel-based alloys (617 and 230) have been selected as structural materials for the advanced gas-cooled reactors, especially for the heat exchanger. An extensive research programme has been launched in France within the framework of the ANTARES programme to evaluate the performances of these materials in VHTR service environment. The experimental work is focused on mechanical properties, thermal stability and corrosion resistance in the temperature range (700-1 000 deg C) over long time. Thus the experimental work includes creep and fatigue tests on as-received materials, short- and medium-term thermal exposure tests followed by tensile and impact toughness tests, short- and medium-term corrosion exposure tests under impure He environment. The status of the results obtained up to now is given in this paper. Additional tests such as long-term thermal ageing and long-term corrosion tests are required to conclude on the selection of the material. (author)

  13. The reactor safety study of experimental multi-purpose VHTR design

    International Nuclear Information System (INIS)

    Yasuno, T.; Mitake, S.; Ezaki, M.; Suzuki, K.

    1981-01-01

    Over the past years, the design works of the Experimental Very High Temperature Reactor (VHTR) plant have been conducted at Japan Atomic Energy Research Institute. The conceptual design has been completed and the more detailed design works and the safety analysis of the experimental VHTR plant are continued. The purposes of design studies are to show the feasibility of the experimental VHTR program, to specify the characteristics and functions of the plant components, to point out the R and D items necessary for the experimental VHTR plant construction, and to analyze the feature of the plant safety. In this paper the summary of system design and safety features of the experimental reactor are indicated. Main issues are the safety philosophy for the design basis accident, the accidents assumed and the engineered safety systems adopted in the design works

  14. An Innovative VHTR Waste Heat Integration with Forward Osmosis Desalination Process

    Energy Technology Data Exchange (ETDEWEB)

    Park, Min Young; Kim, Eung Soo [Seoul National Univ., Seoul (Korea, Republic of)

    2013-10-15

    The integration concept implies the coupling of the waste heat from VHTR with the draw solute recovery system of FO process. By integrating these two novel technologies, advantages, such as improvement of total energy utilization, and production of fresh water using waste heat, can be achieved. In order to thermodynamically analyze the integrated system, the FO process and power conversion system of VHTR are simulated using chemical process software UNISIM together with OLI property package. In this study, the thermodynamic analysis on the VHTR and FO integrated system has been carried out to assess the feasibility of the concept. The FO process including draw solute recovery system is calculated to have a higher GOR compared to the MSF and MED when reasonable FO performance can be promised. Furthermore, when FO process is integrated with the VHTR to produce potable water from waste heat, it still shows a comparable GOR to typical GOR values of MSF and MED. And the waste heat utilization is significantly higher in FO than in MED and MSF. This results in much higher water production when integrated to the same VHTR plant. Therefore, it can be concluded that the suggested integrated system of VHTR and FO is a very promising and strong system concept which has a number of advantages over conventional technologies.

  15. An Assessment of the VHTR Safety Distance Using the Reliability Physics Model

    International Nuclear Information System (INIS)

    Lee, Joeun; Kim, Jintae; Jae, Moosung

    2015-01-01

    In Korea planning the production of hydrogen using high temperature from nuclear power is in progress. To produce hydrogen from nuclear plants, supplying temperature above 800 .deg. C is required. Therefore, Very High Temperature Reactor (VHTR) which is able to provide about 950 .deg. C is suitable. In situation of high temperature and corrosion where hydrogen might be released easily, hydrogen production facility using VHTR has a danger of explosion. Moreover explosion not only has a bad influence upon facility itself but also on VHTR. Those explosions result in unsafe situation that cause serious damage. However, In terms of thermal-hydraulics view, long distance makes low efficiency Thus, in this study, a methodology for the safety assessment of safety distance between the hydrogen production facilities and the VHTR is developed with reliability physics model. Based on the standard safety criteria which is a value of 1 x 10 -6 , the safety distance between the hydrogen production facilities and the VHTR using reliability physics model are calculated to be a value of 60m - 100m. In the future, assessment for characteristic of VHTR, the capacity to resist pressure from outside hydrogen explosion and the overpressure for the large amount of detonation volume in detail is expected to identify more precise safety distance using this reliability physics model

  16. Building competencies for New Gen IV Reactors

    International Nuclear Information System (INIS)

    Pavel, G.L.; Ghitescu, P.

    2015-01-01

    The Advanced Lead Fast Reactor European Demonstrator - ALFRED is designed and sustained by several European countries. It is a 300 MWt (125 MWe) reactor, intended to be built in Romania, near the Pitesti site. Pure lead is used as primary coolant and it is foreseen to have a 40% thermal efficiency. Secondary cycle contains superheated water steam at around 450 Celsius degrees. Through ARCADIA cooperation, 26 partners from all over Europe joined their forces to provide the necessary research support for ALFRED. In Romania, several entities are providing nuclear courses but only the University Politechnica of Bucharest is offering a complete training program for nuclear industry but targeted courses for LFR technology need to be developed and implemented. Issues like physics of breeding, coolant analysis and behavior, targeted computer codes, core design and dynamics, safety still needs to be tackled

  17. Numerical investigation of the High Temperature Reactor (VHTR) using computational fluid dynamics; Investigacao numerica do Reator de Alta Temperatura (VHTR) utilizando fluidodinamica computacional

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Joao Pedro C.T.A.; Santos, Andre A. Campagnole dos; Mesquita, Amir Z., E-mail: jpctap@cdtn.br, E-mail: aacs@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG),Belo Horizonte, MG (Brazil). Lab. de Termo-Hidraulica

    2013-07-01

    This work consists to evaluate and continue the study that is being developed in the Laboratory of Thermo-Hydraulics of the CNEN/CDTN (Centro de Desenvolvimento da Tecnologia Nuclear), aiming to validate the methods and procedures used in the numerical calculations of fluid flow in fuel elements of the core of the VHTR.

  18. WAVE PROPAGATION in the HOT DUCT of VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz; Jim C. P. Liou

    2013-07-01

    In VHTR, helium from the reactor vessel is conveyed to a power conversion unit through a hot duct. In a hypothesized Depressurized Conduction Cooldown event where a rupture of the hot duct occurs, pressure waves will be initiated and reverberate in the hot duct. A numerical model is developed to quantify the transients and the helium mass flux through the rupture for such events. The flow path of the helium forms a closed loop but only the hot duct is modeled in this study. The lower plum of the reactor vessel and the steam generator are treated as specified pressure and/or temperature boundary to the hot duct. The model is based on the conservation principles of mass, momentum and energy, and on the equations of state for helium. The numerical solution is based on the method of characteristics with specified time intervals with a predictor and corrector algorithm. The rupture sub-model gives reasonable results. Transients induced by ruptures with break area equaling 20%, 10%, and 5% of the duct cross-sectional area are described.

  19. Indicial response test for the support post structure of VHTR

    International Nuclear Information System (INIS)

    Futakawa, Masatoshi; Kikuchi, Kenji; Tachibana, Katsumi; Muto, Yasushi

    1985-11-01

    Fuel blocks and removable reflector blocks, which constitute a core of VHTR, are supported by support posts. Each support post is in contact with a hot plenum block at the top end and with a lower plenum block at the bottom end through hemispherical seats to absorb a relative displacement generated by the lateral movement of both blocks by means of small inclination or rotation of support posts. Indicial response tests have been carried out by using a specified one-dimensional vibration model in order to estimate the effects of the support post length, the mass of hot plenum block and the hemispherical radii of both support and post seat on the vibrational characteristics in the support post structure. Futhermore the experimental results have been compared with the analytical ones obtained from the Lagrange's equation. The following are the conclusions derived. (1) The hemispherical radii of support post and post seat have a large effect on the frequency of vibration in the support post structure. (2) The frequency of vibration in the support post structure is predictable using the Lagrange's equation. (author)

  20. TRISO fuel thermal simulations in the LS-VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, Mario C.; Scari, Maria E.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F., E-mail: marc5663@gmail.com, E-mail: melizabethscari@yahoo.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2017-07-01

    The liquid-salt-cooled very high-temperature reactor (LS-VHTR) is a reactor that presents very good characteristics in terms of energy production and safety aspects. It uses as fuel the TRISO particles immersed in a graphite matrix with a cylindrical shape called fuel compact, as moderator graphite and as coolant liquid salt Li{sub 2}BeF{sub 4} called Flibe. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core by performing different simplifications to represent the reactor core and the fuel compact under steady-state conditions, starting the modeling from a single fuel element, until complete the studies with the entire core model developed in the RELAP5-3D code. Two models were considered for representation of the fuel compact, homogeneous and non-homogeneous models, as well as different geometries of the heat structures was considered. The aim to develop several models was to compare the thermal hydraulic characteristics resulting from the construction of a more economical and less discretized model with much more refined models that can lead to more complexes analyzes to representing TRISO effect particles in the fuel compact. The different results found, mainly, for the core temperature distributions are presented and discussed. (author)

  1. Current Status of the Elevated Temperature Structure Design Codes for VHTR

    International Nuclear Information System (INIS)

    Kim, Jong-Bum; Kim, Seok-Hoon; Park, Keun-Bae; Lee, Won-Jae

    2006-01-01

    An elevated temperature structure design and analysis is one of the key issues in the VHTR (Very High Temperature Reactor) project to achieve an economic production of hydrogen which will be an essential energy source for the near future. Since the operating temperature of a VHTR is above 850 .deg. C, the existing code and standards are insufficient for a high temperature structure design. Thus the issues concerning a material selection and behaviors are being studied for the main structural components of a VHTR in leading countries such as US, France, UK, and Japan. In this study, the current status of the ASME code, French RCC-MR, UK R5, and Japanese code were investigated and the necessary R and D items were discussed

  2. Preliminary Estimation of Local Bypass Flow Gap Sizes for a Prismatic VHTR Core

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Jo, Chang Keun; Lee, Won Jae

    2009-01-01

    The Very High Temperature Reactor (VHTR) has been selected for the Nuclear Hydrogen Development and Demonstration (NHDD) project. In the VHTR design, core bypass flow has been one of key issues for core thermal margins and target temperature of the core outlet. The core bypass flow in the prismatic VHTR varies with the core life due to the irradiation shrinkage/ swelling and thermal expansion of the graphite blocks, which could be a significant proportion of the total core flow. Thus, accurate prediction of the bypass flow is of major importance in assuring the core thermal margin. To predict the bypass flow, first of all, local gap sizes between graphite blocks in the core should be determined. The objectives of this work are to develop a methodology for determining the gap sizes and to perform a preliminary evaluation for a reference reactor

  3. An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

    International Nuclear Information System (INIS)

    Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

    2006-01-01

    KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

  4. Nuclear analysis of the experimental VHTR fuel lattice

    International Nuclear Information System (INIS)

    Doi, Takeshi; Shindo, Ryuiti; Hirano, Mitsumasa; Takano, Makoto

    1984-11-01

    Nuclear properties of a fuel lattice in the experimental VHTR core were analyzed with DELIGHT-6 and SRAC codes. Analytical results by both codes were compared by using various calculational model. The nuclear parameters were analyzed, such as a multiplication factor of a fuel lattice and it's variation with burnup, a temperature effect on reactivity, an effect of double-heterogeniety in a resonance absorption calculation, a resonance integral of 238 U and a reactivity worth of burnable poison. From these analyses, following results were obtained. Firstly, on calculational models, 1) Effect of double-heterogeniety in the resonance absorption calculation for Mark-III fuel element, causing a decrease of about 5.5 barns in the resonance integral and an increase of about 2.6 %ΔK in the infinite multiplication factor, 2) The heterogeneous calculation with the collision probability method resulted in about 0.6 %ΔK higher the multiplication factor of fuel lattice than that with the point model, 3) The reactivity worth of burnable poison rod by a multi-region model is about 20 % less than that by a 2-region model at an initial state of burnup and it's variation with burnup are fairly different, Secondly, on comparison between the results by DELIGHT-6 and SRAC, 4) The nuclear parameters obtained with both codes agreed well, Lastly, on the improvement in DELIGHT-6, 5) Consideration of the neutron spectrum shielding effect in the resonance effective cross section calculation caused a decrease of about 2.4 %ΔK in the multiplication factor of fuel lattice, 6) The lattice multiplication factor increased about 0.5 %ΔK by introducing lambda-parameters for the non-resonant nuclie. (J.P.N.)

  5. Guidelines for the structural design of experimental multi-purpose VHTR at the elevated temperature services

    International Nuclear Information System (INIS)

    Nomura, Sueo; Uga, Takeo; Miyamoto, Yoshiaki; Muto, Yasushi; Ikushima, Takeshi

    1976-02-01

    The guidelines are presented for structural design of the experimental multi-purpose VHTR(Very High Temperature Reactor) at the elevated temperature services. Covered are features of the VHTR structural design, specifications, safety design, seismic design, failure modes to be considered, stress criteria for various load combinations and the mechanical properties of the materials. The guidelines were prepared by referring to safety criteria of high-temperature gas cooled reactors, ASME Boiler and Pressure Vessel code, Section III, case 1592 and the domestic seismic design guide of nuclear power facilities. (auth.)

  6. GenBank

    Data.gov (United States)

    U.S. Department of Health & Human Services — GenBank is the NIH genetic sequence database, an annotated collection of all publicly available DNA sequences. GenBank is designed to provide and encourage access...

  7. Study on the tritium behaviors in the VHTR system. Part 2: Analyses on the tritium behaviors in the VHTR/HTSE system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eung S. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States); Oh, Chang H., E-mail: Chang.Oh@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States); Patterson, Mike [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States)

    2010-07-15

    Tritium behaviors in the very high temperature gas reactor (VHTR)/high temperature steam electrolysis (HTSE) system have been analyzed by the TPAC developed by Idaho National Laboratory (INL). The reference system design and conditions were based on the indirect parallel configuration between a VHTR and a HTSE. The analyses were based on the SOBOL method, a modern uncertainty and sensitivity analyses method using variance decomposition and Monte Carlo method. A total of 14 parameters have been taken into account associated with tritium sources, heat exchangers, purification systems, and temperatures. Two sensitivity indices (first order index and total index) were considered, and 15,360 samples were totally used for solution convergence. As a result, important parameters that affect tritium concentration in the hydrogen product have been identified and quantified with the rankings. Several guidelines and recommendations for reducing modeling uncertainties have been also provided throughout the discussions along with some useful ideas for mitigating tritium contaminations in the hydrogen product.

  8. Value-creating investment strategies to manage risk from structural market uncertainties: Switching and compound options in (V)HTR technologies - HTR2008-58157

    International Nuclear Information System (INIS)

    Lauferts, U.; Halbe, C.; Van Heek, A.

    2008-01-01

    To measure the value of a technology investment under uncertainty with standard techniques like net present value (NPV) or return on investment (ROI) will often uncover the difficulty to present convincing business case. Projected cash flows are inefficient or the discount rate chosen to compensate for the risk is so high, that it is disagreeable to the investor s requirements. Decision making and feasibility studies have to look beyond traditional analysis to reveal the strategic value of a technology investment. Here, a Real Option Analysis (ROA) offers a powerful alternative to standard discounted cash-flow (DCF) methodology by risk-adjusting the cash flow along the decision path rather than risk adjusting the discount rate. Within the GEN IV initiative attention is brought not only towards better sustainability, but also to broader industrial application and improved financing. Especially the HTR design is full of strategic optionalities: The high temperature output facilitates penetration into other non-electricity energy markets like industrial process heat applications and the hydrogen market. The flexibility to switch output in markets with multi-source uncertainties reduces downside risk and creates an additional value of over 50% with regard to the Net Present Value without flexibility. The supplement value of deploying a modular (V)HTR design adds over 100% to the project value using real option evaluation tools. Focus of this paper was to quantify the strategic value that comes along a) with the modular design; a design that offers managerial flexibility adapting a step-by-step investment strategy to the actual market demand and b) with the option to switch between two modes of operation, namely electricity and hydrogen production. We will demonstrate that the effect of uncertain electricity prices can be dampened down with a modular HTR design. By using a real option approach, we view the project as a series of compound options - each option depending

  9. Liquid Salt Heat Exchanger Technology for VHTR Based Applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

    2012-10-11

    The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

  10. Numerical investigation of the High Temperature Reactor (VHTR) using computational fluid dynamics

    International Nuclear Information System (INIS)

    Pinto, Joao Pedro C.T.A.; Santos, Andre A. Campagnole dos; Mesquita, Amir Z.

    2013-01-01

    This work consists to evaluate and continue the study that is being developed in the Laboratory of Thermo-Hydraulics of the CNEN/CDTN (Centro de Desenvolvimento da Tecnologia Nuclear), aiming to validate the methods and procedures used in the numerical calculations of fluid flow in fuel elements of the core of the VHTR

  11. Overview of the Modified SI Cycle to Produce Nuclear Hydrogen Coupled to VHTR

    International Nuclear Information System (INIS)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

    2016-01-01

    The steam reforming of methane is one of hydrogen production processes that rely on cheap fossil feedstocks. An overview of the VHTR-based nuclear hydrogen production process with the modified SI cycle has been carried out to establish whether it can be adopted as a feasible technology to produce nuclear hydrogen

  12. A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2005-02-01

    In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

  13. Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving VHTR Efficiency and Testing Material Compatibility - Final Report

    International Nuclear Information System (INIS)

    Chang H. Oh

    2006-01-01

    Generation IV reactors will need to be intrinsically safe, having a proliferation-resistant fuel cycle and several advantages relative to existing light water reactor (LWR). They, however, must still overcome certain technical issues and the cost barrier before it can be built in the U.S. The establishment of a nuclear power cost goal of 3.3 cents/kWh is desirable in order to compete with fossil combined-cycle, gas turbine power generation. This goal requires approximately a 30 percent reduction in power cost for state-of-the-art nuclear plants. It has been demonstrated that this large cost differential can be overcome only by technology improvements that lead to a combination of better efficiency and more compatible reactor materials. The objectives of this research are (1) to develop a supercritical carbon dioxide Brayton cycle in the secondary power conversion side that can be applied to the Very-High-Temperature Gas-Cooled Reactor (VHTR), (2) to improve the plant net efficiency by using the carbon dioxide Brayton cycle, and (3) to test material compatibility at high temperatures and pressures. The reduced volumetric flow rate of carbon dioxide due to higher density compared to helium will reduce compression work, which eventually increase plant net efficiency

  14. Initial VHTR accident scenario classification: models and data.

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Feldman, E. E.; Pointer, W. D.; Wei, T. Y. C.; Nuclear Engineering Division

    2005-09-30

    mixed convection regime for circular channel geometry were identified in the literature. We describe the use of computational experiments to obtain correction factors for applying these circular channel results to the specialized channel geometry of the RCCS. The intent is to reduce the number of laboratory experiments required. The FLUENT and Star-CD codes contain models that in principle can handle mixed convection but no data were found to indicate that their empirical models for turbulence have been benchmarked for mixed convection conditions. Separate effects experiments were proposed for gathering the needed data. In future work we will use the PIRTs to guide review of other components and phenomena in a similar manner as was done for the mixed convection mode in the RCCS. This is consistent with the project objective of identifying weaknesses or gaps in the code models for representing thermal-hydraulic phenomena expected to occur in the VHTR both during normal operation and upsets, identifying the models that need to be developed, and identifying the experiments that must be performed to support model development.

  15. GenBank

    OpenAIRE

    Benson, Dennis A.; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Rapp, Barbara A.; Wheeler, David L.

    2002-01-01

    The GenBank sequence database incorporates publicly available DNA sequences of more than 105 000 different organisms, primarily through direct submission of sequence data from individual laboratories and large-scale sequencing projects. Most submissions are made using the BankIt (web) or Sequin programs and accession numbers are assigned by GenBank staff upon receipt. Data exchange with the EMBL Data Library and the DNA Data Bank of Japan helps ensure comprehensive worldwide coverage. GenBank...

  16. A Review on the VHTR PIRT Development Status of Both Regulatory Authority and Licensee

    International Nuclear Information System (INIS)

    Hwang, Su Hyun; Jeon, Seong Su; Hong, Soon Joon; Lee, Byung Chul; Huh, Chang Wook; Jin, Chang Yong; Kim, Kyun Tae

    2011-01-01

    The VHTR (Very High Temperature Reactor) is defined as a helium-cooled, graphite moderated reactor with a core outlet temperature in excess of 900 .deg. C and a long-term goal of achieving an outlet temperature of 1000 .deg. C. The VHTR is suited for a broad range of applications, including the production of hydrogen and electricity. The PIRT (Phenomena Identification and Ranking Table) provides a structured means of identifying and analyzing a wide variety of off-normal sequences that potentially challenge the viability of complex technological systems. As applied to VHTR, the PIRT is used to identify a spectrum of safety-related sequences or phenomena that could affect those systems, and to rank order those sequences on the basis of their frequencies, their potential consequences, and state of knowledge related to associate phenomena. It is to be used as an early screening tool to identify, categorize, and characterize phenomena and issues that are potentially important to risk and safety of VHTR. Since a specific design has not yet been selected for the choice of the US VHTR (NGNP), it was decided early on to focus on a generic plant and reactor design with broadly typical features. Both a generic Pebble Bed Reactor (PBR) design and a generic Prismatic Modular Reactor (PMR) design were selected as the reference plant for KAERI and ANL PIRT. The generic PBR design selected is a version of the 400 MWt South African PBMR design. The generic PMR design selected is a version of the 600 MWt GT-MHR. The reference plant of NRC PIRT is assumed to be a modular high temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GT-MHR) version (a prismatic-core modular reactor- PMR) or a pebble bed modular reactor (PBMR) version (a pebble bed reactor-PBR) design, with either a director indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. The difference of VHTR PIRT

  17. Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

  18. Studies on the core-support carbon material for VHTR, (1)

    International Nuclear Information System (INIS)

    Matsuo, Hideto; Saito, Tamotsu; Fukuda, Yasumasa; Sasaki, Yasuichi; Hasegawa, Takashi.

    1979-11-01

    To obtain information of core-support carbon material for VHTR, thermal conductivity and electrical resistivity of three domestic carbon blocks were measured. Results indicated the need for development of carbon material with lower thermal conductivity for VHTR. These two were also measured of the samples heat-treated between 1000 0 C and 3040 0 C for one hour. Thermal conductivity increased with heat-treatment above 1200 0 C and resistivity stayed constant between 1500 0 C and 2000 0 C. The results should be useful in choosing the final heat-treatment temperature in carbon material production. The changes of Lorentz number with heat treatment were classified into three heat-treatment temperature regions of below 1500 0 C, 1500 0 C - 2500 0 C, and above 2500 0 C; the results are interpreted with a graphitization model. (author)

  19. A study on bypass flow gap distribution in a prismatic VHTR core

    International Nuclear Information System (INIS)

    Kim, M. H.; Jo, C. K.; Lim, H. S.

    2010-01-01

    Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of irradiation fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass flow and the location of core hot spots are closely related and a measure to reduce the bypass flow is necessary. (authors)

  20. A System Structure for a VHTR-SI Process Dynamic Simulation Code

    International Nuclear Information System (INIS)

    Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung

    2008-01-01

    The VHTR-SI process dynamic simulation code embedded in a mathematical solution engine is an application software system that simulates the dynamic behavior of the VHTR-SI process. Also, the software system supports a user friendly graphical user interface (GUI) for user input/out. Structured analysis techniques were developed in the late 1970s by Yourdon, DeMarco, Gane and Sarson for applying a systematic approach to a systems analysis. It included the use of data flow diagrams and data modeling and fostered the use of an implementation-independent graphical notation for a documentation. In this paper, we present a system structure for a VHRT-SI process dynamic simulation code by using the methodologies of structured analysis

  1. Design studies for the Mark-III core of experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

    1979-08-01

    The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

  2. Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

    Energy Technology Data Exchange (ETDEWEB)

    Glenn E McCreery; Keith G Condie

    2006-09-01

    The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. The present document addresses experimental modeling of flow and thermal mixing phenomena of importance during normal or reduced power operation and during a loss of forced reactor cooling (pressurized conduction cooldown) scenario. The objectives of the experiments are, 1), provide benchmark data for assessment and improvement of codes proposed for NGNP designs and safety studies, and, 2), obtain a better understanding of related phenomena, behavior and needs. Physical models of VHTR vessel upper and lower plenums which use various working fluids to scale phenomena of interest are described. The models may be used to both simulate natural convection conditions during pressurized conduction cooldown and turbulent lower plenum flow during normal or reduced power operation.

  3. Neutronic performance of high molecular weight coolants for a prismatic VHTR

    International Nuclear Information System (INIS)

    Schriener, T. M.; El-Genk, M. S.

    2008-01-01

    A neutronic model is developed of a prismatic Very High Temperature Reactor (VHTR) to investigate the effects on the excess reactivity and operation cycle length of replacing helium with binary gas mixtures of He-Ne, He-N 2 , or He-Xe as reactor coolants and working fluids in the direct Closed Brayton Cycle (CBC) for energy conversion. Also investigated is the neutron activation of these binary gas mixtures in the VHTR. The motivation for using the heavy binary mixtures is the smaller size and the fewer number of stages of the CBC turbo-machinery. The present analysis uses the Monte Carlo code MCNPX 2.6D at typical operating conditions (500-1000 degrees and 7.12 MPa) in the VHTR. He-Ne (15 g/mol) is the best neutronically, but not thermal-hydraulically, followed by He-N 2 . Although He-Ne has ∼13.6% lower heat transfer coefficient than helium, it insignificantly affects the initial excess reactivity and the operation life cycle and experiences no neutrons activation. On the other hand, He-N 2 has 4.4% higher heat transfer coefficient than helium and experiences insignificant neutron activation in the reactor, but decreases the initial excess reactivity by ∼5.2% and the operation cycle length by 6.7%. He-Xe (15 g/mol) has 8% higher heat transfer coefficient than helium, but decreases the initial excess reactivity by 18.2% and the operational cycle length by 17%. In addition, neutron activation of xenon produces a significant source term, requiring shielding of the CBC loop and could contaminate the turbo-machinery with long-lived radioactive cesium. Thus, He-Xe is not recommended as a reactor coolant, but could be used as working fluid in a CBC loop that is indirectly coupled to helium cooled VHTR. (authors)

  4. Reference design (MK-I and MK-II) for experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Miyamoto, Yoshiaki; Suzuki, Kunihiko; Sato, Sadao

    1975-10-01

    This report summarizes the results of a study on thermal and mechanical performances of the core, which are obtained in course of reference design (Mk-I and Mk-II) for the experimental multi-purpose VHTR: (1) Design criteria, design methods and design data. These bases are also discussed in order to refer in the case of proceeding a next design work. (2) The results of performance analysis such as the initial core and its prediction for the irradiated core. (auth.)

  5. Application of Looped Network Analysis Method to Core of Prismatic VHTR

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

    2016-01-01

    Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively as shown in Fig. 1. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Moreover, it is hard to cover whole cases corresponding to the various bypass gap distribution in the whole VHTR core. In order to solve this problem, in this study, the flow network analysis code, FastNet (Flow Analysis for Steady-state Network), was developed using the Looped Network Analysis Method. The applied method was validated by comparing with SNU VHTR multi-block experiment. A 3-demensional network modeling was conducted representing flow paths as flow resistances. Flow network analysis code, FastNet, was developed to evaluate the core bypass flow distribution by using looped network analysis method. Complex flow network could be solved simply by converting the non-linear momentum equation to the linearized equation. The FastNet code predicted the flow distribution of the SNU multi-block experiment accurately

  6. A study on a reliability assessment methodology for the VHTR safety systems

    International Nuclear Information System (INIS)

    Lee, Hyung Sok

    2012-02-01

    The passive safety system of a 300MWt VHTR (Very High Temperature Reactor)which has attracted worldwide attention recently is actively considered for designing the improvement in the safety of the next generation nuclear power plant. The passive system functionality does not rely on an external source of the electrical support system,but on an intelligent use of the natural phenomena, such as convection, conduction, radiation, and gravity. It is not easy to evaluate quantitatively the reliability of the passive safety for the risk analysis considering the existing active system failure since the classical reliability assessment method could not be applicable. Therefore a new reliability methodology needs to be developed and applied for evaluating the reliability of the conceptual designed VHTR in this study. The preliminary evaluation and conceptualization are performed using the concept of the load and capacity theory related to the reliability physics model. The method of response surface method (RSM) is also utilized for evaluating the maximum temperature of nuclear fuel in this study. The significant variables and their correlation are considered for utilizing the GAMMA+ code. The proposed method might contribute to designing the new passive system of the VHTR

  7. Monte Carlo simulation of VHTR particle fuel with chord length sampling

    International Nuclear Information System (INIS)

    Ji, W.; Martin, W. R.

    2007-01-01

    The Very High Temperature Gas-Cooled Reactor (VHTR) poses a problem for neutronic analysis due to the double heterogeneity posed by the particle fuel and either the fuel compacts in the case of the prismatic block reactor or the fuel pebbles in the case of the pebble bed reactor. Direct Monte Carlo simulation has been used in recent years to analyze these VHTR configurations but is computationally challenged when space dependent phenomena are considered such as depletion or temperature feedback. As an alternative approach, we have considered chord length sampling to reduce the computational burden of the Monte Carlo simulation. We have improved on an existing method called 'limited chord length sampling' and have used it to analyze stochastic media representative of either pebble bed or prismatic VHTR fuel geometries. Based on the assumption that the PDF had an exponential form, a theoretical chord length distribution is derived and shown to be an excellent model for a wide range of packing fractions. This chord length PDF was then used to analyze a stochastic medium that was constructed using the RSA (Random Sequential Addition) algorithm and the results were compared to a benchmark Monte Carlo simulation of the actual stochastic geometry. The results are promising and suggest that the theoretical chord length PDF can be used instead of a full Monte Carlo random walk simulation in the stochastic medium, saving orders of magnitude in computational time (and memory demand) to perform the simulation. (authors)

  8. State of the Art Report for a Bearing for VHTR Helium Circulator

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Song, Kee Nam; Kim, Yong Wan; Lee, Won Jae

    2008-10-01

    A helium circulator in a VHTR(Very High Temperature gas-cooled Reactor) plays a core role which translates thermal energy at high temperature from a nuclear core to a steam generator. Helium as a operating coolant circulates a primary circuit in high temperature and high pressure state, and controls thermal output of a nuclear core by controlling flow rate. A helium circulator is the only rotating machinery in a VHTR, and its reliability should be guaranteed for reliable operation of a reactor and stable production of hydrogen. Generally a main helium circulator is installed on the top of a steam generator vessel, and helium is circulated only by a main helium circulator in a normal operation state. An auxiliary or shutdown circulator is installed at the bottom of a reactor vessel, and it is an auxiliary circulator for shutting down a reactor in case of refueling or accelerating cooling down in case of fast cooling. Since a rotating shaft of a helium circulator is supported by bearings, bearings are the important machine elements which determines reliability of a helium circulator and a nuclear reactor. Various types of support bearings have been developed and applied for circulator bearings since 1960s, and it is still developing for developing VHTRs. So it is necessary to review and analyze the current technical state of helium circulator support bearings to develop bearings for Koran developing VHTR helium circulator

  9. Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

    Energy Technology Data Exchange (ETDEWEB)

    Fok, Alex

    2013-10-30

    The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

  10. Development of flow network analysis code for block type VHTR core by linear theory method

    International Nuclear Information System (INIS)

    Lee, J. H.; Yoon, S. J.; Park, J. W.; Park, G. C.

    2012-01-01

    VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

  11. Preliminary Core Design Analysis of a 200MWth Pebble Bed-type VHTR

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Noh, Jae Man

    2007-01-01

    This paper intends to suggest the preliminary core design analysis of a VHTR for a hydrogen production. The nuclear hydrogen system that utilizes the high temperature heat generated from the VHTR is a promising candidate for a cost effective, safe and clean supply of hydrogen in the age of hydrogen economy. Among two candidate VHTR cores, that is, a prismatic modular reactor (PMR) and a pebble bed-type reactor (PBR), we focus on the design of a 200MWth PBR (hereinafter PBR200) in this paper. Here, the 200MWth power is selected for a demonstration plant. The core configuration of the PBR200 is similar to the PBMR (Pebble Bed Modular Reactor, 400MWth) of South Africa, but the overall dimension of the reactor system is scaled-down. This paper is to suggest two candidate PBR200 cores. One is an annular core with an inner reflector (PBR200-CD1) which was presented at IWRES07, and the other is a cylindrical core without an inner reflector (PBR200-CD2)

  12. Effect of Permanent Side Reflector on the Temperature Variation in the VHTR Core

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Nam; Tak, Nam-il; Kim, Min-Hwan [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The temperature and pressure conditions range from 490°C to 950°C, 7MPa. GAMMA+ was developed to predict the overall phenomena of the VHTR system. The GAMMA+ algorithms focused on the transient condition for the systems. Therefore, the computational control volumes are coarse for reducing the computational time. However, there are difficulties calculating the temperature gradient in the fuel blocks in detail. There is a demand to predict a hot spot and temperature distribution in the reactor core to apply a thermal stress and find the fuel temperature margin. Computational Fluid Dynamic (CFD) tools can be an option to model the VHTR. However, the fluid has to be solved in three dimensions. The long computational time and heavy burden of the memory size have called for an alternative option. The PSR blocks are considered in the prismatic VHTR calculation with the CORONA code. The temperatures of a single assembly with an arc shape reflector by the CORONA code were verified with the results by the CFX calculation. The temperature distributions of the PSR regions did not show significant differences depending on the fixed inlet temperature boundary condition and bypass flow condition.

  13. Uncertainty analysis methods for estimation of reliability of passive system of VHTR

    International Nuclear Information System (INIS)

    Han, S.J.

    2012-01-01

    An estimation of reliability of passive system for the probabilistic safety assessment (PSA) of a very high temperature reactor (VHTR) is under development in Korea. The essential approach of this estimation is to measure the uncertainty of the system performance under a specific accident condition. The uncertainty propagation approach according to the simulation of phenomenological models (computer codes) is adopted as a typical method to estimate the uncertainty for this purpose. This presentation introduced the uncertainty propagation and discussed the related issues focusing on the propagation object and its surrogates. To achieve a sufficient level of depth of uncertainty results, the applicability of the propagation should be carefully reviewed. For an example study, Latin-hypercube sampling (LHS) method as a direct propagation was tested for a specific accident sequence of VHTR. The reactor cavity cooling system (RCCS) developed by KAERI was considered for this example study. This is an air-cooled type passive system that has no active components for its operation. The accident sequence is a low pressure conduction cooling (LPCC) accident that is considered as a design basis accident for the safety design of VHTR. This sequence is due to a large failure of the pressure boundary of the reactor system such as a guillotine break of coolant pipe lines. The presentation discussed the obtained insights (benefit and weakness) to apply an estimation of reliability of passive system

  14. Evaluation of Core Bypass Flow in the Prismatic VHTR with a Multi-block Experiment

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Kim, Min Hwan

    2010-01-01

    The core of Prismatic Modular Reactor (PMR) consists of assemblies of hexagonal graphite fuel and reflector elements. The core bypass flow of Very High Temperature Reactor (VHTR) is defined as the core flow that does not pass through the coolant channels but passes through the bypass gap between fuel elements. The increase in bypass flow makes the decrease in effective coolant flow. Since the core bypass flow has a negative impact on safety and efficiency of VHTR, core bypass phenomena have to be investigated to improve the core thermal margin of VHTR. For this purpose, the international project, I-NERI project, has been carried out since 2008. I-NERI project is collaborative project that KAERI and SNU of Korea side and INL, ANL and TAMU of U.S side are involved. In order to evaluate the core bypass flow, the multicolumn and multi-layer experimental facility is designed by SNU. In this experiment, the effect of cross-flow and local variation of bypass gap on the bypass flow distribution is investigated. Furthermore, the experimental data will be used for validation of CFD code or thermal hydraulic analysis codes such as GAMMA or GAS-NET

  15. CFD Validation with a Multi-Block Experiment to Evaluate the Core Bypass Flow in VHTR

    International Nuclear Information System (INIS)

    Yoon, Su Jong; Lee, Jeong Hun; Park, Goon Cherl; Kim, Min Hwan

    2010-01-01

    Core bypass flow of Very High Temperature Reactor (VHTR) is defined as the ineffective coolant which passes through the bypass gaps between the block columns and the crossflow gaps between the stacked blocks. This flows lead to the variation of the flow distribution in the core and affect the core thermal margin and the safety of VHTR. Therefore, bypass flow should be investigated and quantified. However, it is not a simple question, because the flow path of VHTR core is very complex. In particular, since dimensions of the bypass gap and the crossflow gap are of the order of few millimeters, it is very difficult to measure and to analyze the flow field at those gaps. Seoul National University (SNU) multi-block experiment was carried out to evaluate the bypass flow distribution and the flow characteristics. The coolant flow rate through outlet of each block column was measured, but the local flow field was measured restrictively in the experiment. Instead, CFD analysis was carried out to investigate the local phenomena of the experiment. A commercial CFD code CFX-12 was validated by comparing the simulation results and the experimental data

  16. Safety study of the coupling of a VHTR with a hydrogen production plant

    International Nuclear Information System (INIS)

    Bertrand, F.; Germain, T.; Bentivoglio, F.; Bonnet, F.; Moyart, Q.; Aujollet, P.

    2011-01-01

    Highlights: → The paper deals with safety issues of the coupling of a VHTR with a H 2 production plant. → Internal incidents/accidents in the coupling system have been studied with the CATHARE2 code. → Transient studies have indicated a substantial grace delay when the VHTR faces the H 2 plant disturbances. → Hydrogen release and combustion leads to safety distances of about 100 m. → No showstopper has been put in evidence regarding the feasibility of the VHTR/H 2 plant coupling. - Abstract: The present paper deals with specific safety issues resulting from the coupling of a nuclear reactor (very high temperature reactor, VHTR) with a hydrogen production plant (HYPP). The first part is devoted to the safety approach consisting in taking into account the safety standards and rules dedicated to the nuclear facility as well as those dedicated to the process industry. This approach enabled two main families of events to be distinguished: the so-called internal events taking place in the coupling circuit (transients, breaks in pipes and in heat exchangers) and the external events able to threat the integrity of the various equipments (in particular the VHTR containment and emergency cooling system) that could result from accidents in the HYPP. By considering a hydrogen production by means of the iodine/sulfur (IS) process, the consequences of the both families of events aforementioned have been assessed in order to provide an order of magnitude of the effects of the incidents and accidents and also in order to propose safety provisions to mitigate these effects when it is necessary. The study of transients induced by a failure of a part of the HYPP has shown the possibility to keep the part of the HYPP unaffected by the transient under operation by means of an adapted regulation set. Moreover, the time to react in case of transfer of corrosive products in the VHTR containment has been assessed as well as the thermohydraulic loading that would experience the

  17. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    International Nuclear Information System (INIS)

    Guzina, Bojan; Kunerth, Dennis

    2014-01-01

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  18. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Guzina, Bojan [Univ. of Minnesota, Minneapolis, MN (United States); Kunerth, Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-30

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  19. GenBank

    OpenAIRE

    Benson, Dennis A.; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Wheeler, David L.

    2006-01-01

    GenBank (R) is a comprehensive database that contains publicly available nucleotide sequences for more than 240 000 named organisms, obtained primarily through submissions from individual laboratories and batch submissions from large-scale sequencing projects. Most submissions are made using the web-based BankIt or standalone Sequin programs and accession numbers are assigned by GenBank staff upon receipt. Daily data exchange with the EMBL Data Library in Europe and the DNA Data Bank of Japan...

  20. GenBank

    OpenAIRE

    Benson, Dennis A.; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Sayers, Eric W.

    2008-01-01

    GenBank? is a comprehensive database that contains publicly available nucleotide sequences for more than 300 000 organisms named at the genus level or lower, obtained primarily through submissions from individual laboratories and batch submissions from large-scale sequencing projects. Most submissions are made using the web-based BankIt or standalone Sequin programs, and accession numbers are assigned by GenBank? staff upon receipt. Daily data exchange with the European Molecular Biology Labo...

  1. GenBank

    OpenAIRE

    Benson, Dennis A.; Cavanaugh, Mark; Clark, Karen; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Sayers, Eric W.

    2012-01-01

    GenBank? (http://www.ncbi.nlm.nih.gov) is a comprehensive database that contains publicly available nucleotide sequences for almost 260 000 formally described species. These sequences are obtained primarily through submissions from individual laboratories and batch submissions from large-scale sequencing projects, including whole-genome shotgun (WGS) and environmental sampling projects. Most submissions are made using the web-based BankIt or standalone Sequin programs, and GenBank staff assig...

  2. Investigation of an Alternative Fuel Form for the Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    Much of the recent studies investigating the use of liquid salts as reactor coolants have utilized a core configuration of graphite prismatic fuel block assemblies with TRISO particles embedded into cylindrical fuel compacts arranged in a triangular pitch lattice. Although many calculations have been performed for this fuel form in gas cooled reactors, it would be instructive to investigate whether an alternative fuel form may yield improved performance for the liquid salt-cooled Very High Temperature Reactor (LS-VHTR). This study investigates how variations in the fuel form will impact the performance of the LS-VHTR during normal and accident conditions and compares the results with a similar analysis that was recently completed for a LS-VHTR core made up of prismatic block fuel. (author)

  3. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  4. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    International Nuclear Information System (INIS)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok

    2016-01-01

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  5. Materials for generation-IV nuclear reactors

    International Nuclear Information System (INIS)

    Alvarez, M. G.

    2009-01-01

    Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

  6. CFD Analysis for Predicting Flow Resistance of the Cross Flow Gap in Prismatic VHTR Core

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Park, Jong Woon

    2011-01-01

    The core of Very High Temperature Reactor (VHTR) consists of assemblies of hexagonal graphite blocks and its height and across-flats width are 800 mm and 360 mm respectively. They are equipped with 108 coolant holes 16 mm in diameter. Up to ten fuel blocks arranged in vertical order form a fuel element column and the neutron flux varies over the cross section of the core. It makes different axial shrinkage of fuel element and this leads to make wedge-shaped gaps between the base and top surfaces of stacked blocks. The cross flow is defined as the core flow that passes through this cross gaps. The cross flow complicates the flow distribution of reactor core. Moreover, the cross flow could lead to uneven coolant distribution and consequently to superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. In particular, to predict amount of flow at the cross flow gap obtaining accurate flow loss coefficient is important. Nevertheless, there has not been much effort in domestic. The experiment of cross flow was carried out by H. G. Groehn in 1981 Germany. For the study of cross flow the applicability of CFD code should be validated. In this paper a commercial CFD code CFX-12 validation will be carried out with this cross flow experiment. Validated data can be used for validation of other thermal-hydraulic analysis codes

  7. VHTR-fuel irradiation capsules for VT-1 hole of JRR-2

    International Nuclear Information System (INIS)

    Kikuchi, Teruo; Kikuchi, Akira; Tobita, Tsutomu; Kashimura, Satoru; Miyasaka, Yasuhiko

    1977-02-01

    Irradiations of VHTR fuels were made in the VT-1 irradiation hole of JRR-2. Three capsules, VP-1, VP-2 and VP-4, which contained fuel compacts, were irradiated for 300 hr at temperatures of 950 0 , 1370 0 and 1500 0 C up to the estimated burn-ups of 0.74, 0.87 and 0.80%FIMA, respectively. And, to study the amoeba effect of fuel particles, two capsules, VP-3 and VP-5, were irradiated for 300 hr at temperatures of 1650 0 and 1670 0 C up to the estimated burn-ups of 0.38 and 0.33%FIMA, respectively. (auth.)

  8. Key technology for (V)HTR: laser beam joining of SiC

    International Nuclear Information System (INIS)

    Knorr, J.; Lippmann, W.; Reinecke, A.M.; Wolf, R.; Rasper, R.; Kerber, A.; Wolter, A.

    2005-01-01

    Laser beam joining has numerous advantages over other methods presently known. After having been developed successful for brazing silicon carbide for high temperature applications, this technology is now also available for silicon nitride. Thus the field of application of SiC and Si 3 N 4 which are very interesting materials for the nuclear sector is considerably extended thanks to this new technology. Ceramic encapsulation of fuel and absorber increases the margins for operation at very high temperatures. Additionally, without ceramic encapsulation of the main core components, it will be difficult to continue claiming non-catastrophic behaviour for the (V)HTR. (orig.)

  9. Improving Corrosion Behavior in SCWR, LFR and VHTR Reactor Materials by Formation of a Stable Oxide

    International Nuclear Information System (INIS)

    Motta, Arthur T.; Comstock, Robert; Li, Ning; Allen, Todd; Was, Gary

    2009-01-01

    The objective of this study is to understand the influence of the alloy microstructure and composition on the formation of a stable, protective oxide in the environments relevant to the SCWR and LFR reactor concepts, as well as to the VHTR. It is proposed to use state-of-the art techniques to study the fine structure of these oxides to identify the structural differences between stable and unstable oxide layers. The techniques to be used are microbeam synchrotron radiation diffraction and fluorescence, and cross-sectional transmission electron microcopy on samples prepared using focused ion beam.

  10. Needs in Research and Development on materials for the gas coolant nuclear system: HTR/VHTR and GFR; Besoins en R et D sur les materiaux pour les systemes nucleaires a caloporteur gaz: HTR/VHTR et GFR

    Energy Technology Data Exchange (ETDEWEB)

    Billot, Ph. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares (DEN/DDIN), 91 - Gif Sur Yvette (France)

    2003-07-01

    This presentation takes stock on the materials for high temperature reactors HTR (850 C), very high temperature VHTR(>1000 C) and fast neutrons high temperature GGF(850 C). It concerns the welding materials for the vessel, Ni-based superalloys for gas turbines, coatings, graphite, ceramics and corrosion studies. (A.L.B.)

  11. Power requirements at the VHTR/HTE interface for hydrogen production

    International Nuclear Information System (INIS)

    Vilim, R.B.

    2007-01-01

    The power requirements at the interface between the High Temperature Electrolysis (HTE) process and the Very High Temperature Reactor (VHTR) were investigated. The study was performed using a network systems code that linked together individual component models for boiler, condenser, turbine, compressor, pump, gas-to-gas heat exchanger, electrolyser, and reactor and properties for water, hydrogen, oxygen, nitrogen, and helium. A species mixture model supported the use of mixtures of gases in each component model. The requirements for a reference design with a dedicated high temperature process heat loop are given. In general the quantity and quality of the process heat needed by the HTE process is a function of how the electrolyser is operated. Operating at higher voltage increases throughput and resistive heating providing the opportunity to recuperate this heat and supplant a large fraction of high temperature reactor heat. Any shortfall can be added by electrical heaters in the HTE plant. Eliminating the associated high temperature heat exchanger from the nuclear plant in this manner would significantly improve safety and maintainability. Low temperature process heat is still needed to vaporize water for the HTE process but this can be obtained at very low cost from VHTR waste heat rejected to the ultimate heat sink. (author)

  12. Numerical investigation of flow characteristics in a prototypical lower plenum of a prismatic VHTR

    International Nuclear Information System (INIS)

    Ying, Alice; Narula, Manmeet; Abdou, Mohamed; Tsai, Peter; Ando, Yuya

    2007-01-01

    The aim of this study is to obtain insights into the flow behavior, as well as to develop predictive capability with regards to the flow and thermal mixing, that occurs in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In this paper, numerical modeling has been used to capture qualitative phenomena observed during an experiment performed at INL, using a finite volume, thermo-fluid solver system, 'SC/Tetra' from CRADLE. The choice of the correct turbulence model is critical to accurately predict the flow in the VHTR lower plenum. Four different turbulence models have been used in this study and the flow predictions are significantly different. A trail of marker particles and fluid temperature as a passive scalar have been used to qualitatively study the flow characteristics, specifically the turbulent mixing of water jets. The quantitative experimental data, when available, will be used to compare and improve on the available turbulence models. Preliminary numerical modeling has been carried out to address the issue of hot streaking and buoyancy effects of hot helium jets in the lower plenum. (author)

  13. Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

    2007-07-15

    Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

  14. Safety assessment of VHTR hydrogen production system against fire, explosion and acute toxicity

    International Nuclear Information System (INIS)

    Murakami, Tomoyuki; Nishihara, Tetsuo; Kunitomi, Kazuhiko

    2008-01-01

    The Japan Atomic Energy Agency has been developing a nuclear hydrogen production system by using heat from the Very High Temperature Reactor (VHTR). This system will handle a large amount of combustible gas and toxic gas. The risk from fire, explosion and acute toxic exposure caused by an accident involving chemical material release in a hydrogen production system is assessed. It is important to ensure the safety of the nuclear plant, and the risks for public health should be sufficiently small. This report provides the basic policy for the safety evaluation in cases of accident involving fire, explosion and toxic material release in a hydrogen production system. Preliminary safety analysis of a commercial-sized VHTR hydrogen production system, GTHTR300C, is performed. This analysis provides us with useful information on the separation distance between a nuclear plant and a hydrogen production system and a prospect that an accident in a hydrogen production system does not significantly increase the risks of the public. (author)

  15. Design studies of back up cores for the experimental multi-purpose VHTR, (1)

    International Nuclear Information System (INIS)

    Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu

    1982-09-01

    For the Experimental Multi-Purpose Very High Temperature Reactor, design studies have been made of two backup cores loaded with new type fuel elements. The purpose is to improve core operational characteristics of the standard design core (Mark-III core) consisting of pin-in-block type fuel element having externally cooled hollow fuel rods. The first backup core (semi-pin fuel core) is composed of fuel elements with internally cooled fuel pins, and the second core (multihole fuel core) is composed of multihole fuel elements, which can be adopted for the experimental VHTR as the substitution of the standard Mark-III fuel element. Either of the cores has 73 fuel columns and 4 m height. The arrangement of active core and reactor internal structure is same as that in the standard design core. These backup cores meet almost all design requirements of the VHTR and increase the margins for some important design items in comparison with the standard core (Mark-III core). This report describes the overall characteristics of nuclear, thermal-hydraulic, fuel and safety, and structural consideration for these cores. (author)

  16. Graphite oxidation and structural strength of graphite support column in VHTR

    International Nuclear Information System (INIS)

    Park, Byung Ha; No, Hee Cheno; Kim, Eung Soo; Oh, Chang H.

    2009-01-01

    The air-ingress event by a large pipe break is an important accident considered in design of very high-temperature gas-cooled reactors (VHTR). Core-collapse prediction is a main safety issue. Structural failure model are technically required. The objective of this study is to develop structural failure model for the supporting graphite material in the lower plenum of the GT-MHR (gas-turbine-modular high temperature reactor). Graphite support column is important for VHTR structural integrity. Graphite support columns are under the axial load. Critical strength of graphite column is related to slenderness ratio and bulk density. Through compression tests for fresh and oxidized graphite columns we show that compressive strength of IG-110 was 79.46 MPa. And, the buckling strength of IG-110 column was expressed by the empirical formula: σ 0 =σ straight-line - C L/r, σ straight-line =91.31 MPa, C=1.01. The results of uniform and non-uniform oxidation tests show that the strength degradation of oxidized graphite column is expressed in the following non-dimensional form: σ/σ 0 =exp(-kd), k=0.111. Also, from the results of the uniform oxidation test with a complicated-shape column, we found out that the above non-dimensional equation obtained from the uniform oxidation test is applicable to a uniform oxidation case with a complicated-shape column. (author)

  17. DETEKSI GEN-GEN PENYANDI FAKTOR VIRULENSI PADA BAKTERI VIBRIO

    Directory of Open Access Journals (Sweden)

    Ince Ayu Khairani Kadriah

    2011-04-01

    menggunakan isolat bakteri yang diisolasi dari budidaya udang windu di berbagai daerah di Sulawesi Selatan dan Jawa. Pada penelitian ini digunakan primer spesifik untuk mendeteksi gen-gen virulen toxR gene, hemolysin (vvh gene, dan GyrB gene dengan metode PCR. Dari 35 isolat yang diisolasi, 20 isolat terdeteksi memiliki gen virulensi dan 8 di antaranya memiliki dua gen virulen. Spesies bakteri yang memiliki gen virulen adalah: V.harveyi, V. parahaemolyticus, V. mimicus, dan V. campbelli

  18. Needs in Research and Development on materials for the gas coolant nuclear system: HTR/VHTR and GFR

    International Nuclear Information System (INIS)

    Billot, Ph.

    2003-01-01

    This presentation takes stock on the materials for high temperature reactors HTR (850 C), very high temperature VHTR(>1000 C) and fast neutrons high temperature GGF(850 C). It concerns the welding materials for the vessel, Ni-based superalloys for gas turbines, coatings, graphite, ceramics and corrosion studies. (A.L.B.)

  19. GenBank.

    OpenAIRE

    Benson, D; Lipman, D J; Ostell, J

    1993-01-01

    The GenBank sequence database has undergone an expansion in data coverage, annotation content and the development of new services for the scientific community. In addition to nucleotide sequences, data from the major protein sequence and structural databases, and from U.S. and European patents is now included in an integrated system. MEDLINE abstracts from published articles describing the sequences provide an important new source of biological annotation for sequence entries. In addition to ...

  20. Improvements of Physical Models in TRITGO code for Tritium Behavior Analysis in VHTR

    International Nuclear Information System (INIS)

    Yoo, Jun Soo; Tak, Nam Il; Lim, Hong Sik

    2010-01-01

    Since tritium is radioactive material with 12.32 year of half-life and is generated by a ternary fission reaction in fuel as well as by neutron absorption reactions of impurities in Very High Temperature gas-cooled Reactor (VHTR) core, accurate prediction of tritium behavior and its concentration in product hydrogen is definitely important in terms of public safety for its construction. In this respect, TRITGO code was developed for estimating the tritium production and distribution in high temperature gas-cooled reactors by General Atomics (GA). However, some models in it are hard-wired to specific reactor type or too simplified, which makes the analysis results less applicable. Thus, major improvements need to be considered for better predictions. In this study, some of model improvements have been suggested and its effect is evaluated based on the analysis work against PMR600 design concept

  1. Evaluation of the oxidation behavior and strength of the graphite components in the VHTR, (1)

    International Nuclear Information System (INIS)

    Eto, Motokuni; Kurosawa, Takeshi; Nomura, Shinzo; Imai, Hisashi

    1987-04-01

    Oxidation experiments have been carried out mainly on a fine-grained isotropic graphite, IG-110, at temperatures between 1173 and 1473 K in a water vapor/helium mixture. In most cases water vapor concentration was 0.65 vol% and helium pressure, 1 atm. Reaction rate and burn-off profile were measured using cylindrical specimens. On the basis of the experimental data the oxidation behavior of fuel block and core support post under the condition of the VHTR operation was estimated using the first-order or Langmuir-Hinshelwood equation with regard to water vapor concentration. Strength and stress-strain relationship of the graphite components with burn-off profiles estimated above were analyzed on the basis of the model for stress-strain relationship and strength of graphite specimens with density gradients. The estimation indicated that the integrity of the components would be maintained during normal reactor operation. (author)

  2. Feasibility study of thermal insulation materials for core support of experimental VHTR

    International Nuclear Information System (INIS)

    Kawakami, H.; Nakanishi, T.

    1982-01-01

    Thermal insulation materials for core support of the experimental VHTR, planned by JAERI, should maintain moderate compressive strength and dimensional stability as well as low thermal conductivity at the maximum service temperature of 1100 0 C for 20 years. For selecting materials, we investigate properties of some candidates, and evaluate their feasibility. Preliminary tests, heat treatment test and compressive creep tests for 1000 hours at 900 0 C and 1000 0 C were conducted. In the preliminary tests, EG-38B (carbon baked at 1350 0 C) and Fine Finnex 600 (silicon nitride) showed acceptable physical stability. In the heat treatment tests, silicon nitride showed weight loss probably caused by thermal decomposition. Compressive creep deformation of Fine Finnex 600 was negligible under stress of 100 kg/cm 2 for 1000 hours. Heat treatment at 1200 to 1300 0 C for 50 hours improved dimensional stability of carbon at 1000 0 C

  3. Analytical study on coolant temperature of several leak flows in the experimental VHTr core

    International Nuclear Information System (INIS)

    Fumizawa, Motoh; Arai, Taketoshi; Miyamoto, Yoshiaki

    1982-08-01

    This report describes heat transfer analysis of several leak flows which bypass main coolant flow path in the experimental VHTR core. The analysis contains the leak flow at permanent reflectors, replaceable reflectors and gaps between fuel columns. The summary of the results are as follows: (1) the temperature of the leak flow gas increases up to the surface temperature of permanent reflectors, (2) the gas temperature at replaceable reflectors increases at least 40 0 C in case of the worst analytical condition, (3) the gas temperature increases remarkably with decreasing equivalent diameter which is changed by the angle of bevel edge of the reflector, (4) while the gas temperature is low at the upper part of the fuel element, the temperature increases rapidly when it flow down along the gap of the fuel columns. (author)

  4. Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

    International Nuclear Information System (INIS)

    Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

    1982-01-01

    A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

  5. A study on the aseismic safety of the experimental VHTR on the dense sandy layer

    International Nuclear Information System (INIS)

    Fujita, Shigeki; Ito, Yoshio; Baba, Osamu; Suzuki, Hideyuki; Takewaki, Naonobu; Kondo, Tsukasa; Yoshimura, Takashi; Yamada, Hitoshi.

    1986-12-01

    A series of studies has been carried out in 1983 and 1985 for the purpose of confirming the aseismic safety of the Experimental VHTR on the dense sandy layer. In 1983, effect of some of soil properties on seismic responses of the reactor building was estimated by means of parametric survey, and soil properties were estimated by analyzing the obserbed earthquake record. In 1985, literature review, linear, nonlinear parametric analyses and nonlinear simulation analyses were carried to study and compare the analysis method. In addition, seismic response of proposed construction site was estimated with nonlinear analysis method. As a result of these studies, the seismic response of reactor building on the dense sandy layers and wave propagation characteristics of sandy layers are understood. Especially, by means of many parametric studies, the effect of input wave characteristics, soil stiffness, nonlinear characteristics of soil properties and nonlinear analysis method on the reactor building responses were evaluated. (author)

  6. Development of Barrier Layers for the Protection of Candidate Alloys in the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Levi, Carlos G. [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States); Jones, J. Wayne [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States); Pollock, Tresa M. [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States); Was, Gary S. [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States)

    2015-01-22

    The objective of this project was to develop concepts for barrier layers that enable leading candi- date Ni alloys to meet the longer term operating temperature and durability requirements of the VHTR. The concepts were based on alpha alumina as a primary surface barrier, underlay by one or more chemically distinct alloy layers that would promote and sustain the formation of the pro- tective scale. The surface layers must possess stable microstructures that provide resistance to oxidation, de-carburization and/or carburization, as well as durability against relevant forms of thermo-mechanical cycling. The system must also have a self-healing ability to allow endurance for long exposure times at temperatures up to 1000°C.

  7. Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Johnson, Richard; Schultz, Richard

    2010-01-01

    The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to better understand bypass flow phenomena and establish an evaluation method for the reactor core using the commercial CFD code FLUENT. Parametric calculations changing several factors in a one-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the resulting flow and temperature distributions in the prismatic core. Factors include inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block and also dramatically increases the variation in coolant channel outlet temperatures for a given block that may have repercussions on the structural integrity of the graphite, the neutronics and the potential for hot streaking and hot spots occurring in the lower plenum.

  8. Creep-fatigue of High Temperature Materials for VHTR: Effect of Cyclic Loading and Environment

    Energy Technology Data Exchange (ETDEWEB)

    Celine Cabet; L. Carroll; R. Wright; R. Madland

    2011-05-01

    Alloy 617 is the one of the leading candidate materials for Intermediate Heat eXchangers (IHX) of a Very High Temperature Reactor (VHTR). System start-ups and shut-downs as well as power transients will produce low cycle fatigue (LCF) loadings of components. Furthermore, the anticipated IHX operating temperature, up to 950°C, is in the range of creep so that creep-fatigue interaction, which can significantly increase the fatigue crack growth, may be one of the primary IHX damage modes. To address the needs for Alloy 617 codification and licensing, a significant creep-fatigue testing program is underway at Idaho National Laboratory. Strain controlled LCF tests including hold times up to 1800s at maximum tensile strain were conducted at total strain range of 0.3% and 0.6% in air at 950°C. Creep-fatigue testing was also performed in a simulated VHTR impure helium coolant for selected experimental conditions. The creep-fatigue tests resulted in failure times up to 1000 hrs. Fatigue resistance was significantly decreased when a hold time was added at peak stress and when the total strain was increased. The fracture mode also changed from transgranular to intergranular with introduction of a tensile hold. Changes in the microstructure were methodically characterized. A combined effect of temperature, cyclic and static loading and environment was evidenced in the targeted operating conditions of the IHX. This paper This paper reviews the data previously published by Carroll and co-workers in references 10 and 11 focusing on the role of inelastic strain accumulation and of oxidation in the initiation and propagation of surface fatigue cracks.

  9. Development of the Log-in Process and the Operation Process for the VHTR-SI Process Dynamic Simulation Code

    International Nuclear Information System (INIS)

    Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung

    2009-01-01

    The VHTR-SI process is a hydrogen production technique by using Sulfur and Iodine. The SI process for a hydrogen production uses a high temperature (about 950 .deg. C) of the He gas which is a cooling material for an energy sources. The Korea Atomic Energy Research Institute Dynamic Simulation Code (KAERI DySCo) is an integration application software that simulates the dynamic behavior of the VHTR-SI process. A dynamic modeling is used to express and model the behavior of the software system over time. The dynamic modeling deals with the control flow of system, the interaction of objects and the order of actions in view of a time and transition by using a sequence diagram and a state transition diagram. In this paper, we present an user log-in process and an operation process for the KAERI DySCo by using a sequence diagram and a state transition diagram

  10. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    International Nuclear Information System (INIS)

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric

    2008-01-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC R process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  11. Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Wang, Lumin; Was, Gary

    2010-01-01

    Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

  12. VHTR-based Nuclear Hydrogen Plant Analysis for Hydrogen Production with SI, HyS, and HTSE Facilities

    International Nuclear Information System (INIS)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

    2016-01-01

    In this paper, analyses of material and heat balances on the SI, HyS, and HTSE processes coupled to a Very High Temperature gas-cooled Reactor (VHTR) were performed. The hydrogen production efficiency including the thermal to electric energy ratio demanded from each process is found and the normalized evaluation results obtained from three processes are compared to each other. The currently technological issues to maintain the long term continuous operation of each process will be discussed at the conference site. VHTR-based nuclear hydrogen plant analysis for hydrogen production with SI, HyS, and HTSE facilities has been carried out to determine the thermal efficiency. It is evident that the thermal to electrical energy ratio demanded from each hydrogen production process is an important parameter to select the adequate process for hydrogen production. To improve the hydrogen production efficiency in the SI process coupled to the VHTR without electrical power generation, the demand of electrical energy in the SI process should be minimized by eliminating an electrodialysis step to break through the azeotrope of the HI/I_2/H_2O ternary aqueous solution

  13. Generation IV reactors: international projects

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

    2003-01-01

    Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

  14. Numerical and experimental investigation on labyrinth seal mechanism for bypass flow reduction in prismatic VHTR core

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su-Jong, E-mail: paper80@snu.ac.r [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Sang-Moon [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Tak, Nam-il; Kim, Min-Hwan [Korea Atomic Energy Research Institute, 150-1 Deokjin-Dong, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of); Kim, Kwang-Yong [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of)

    2013-09-15

    Highlights: • Bypass flow reduction method was developed by applying labyrinth seal mechanism. • Grooves on side walls of replaceable reflector block were made. • Design of the grooved wall of the reflector block was optimized by the RSA method. • The flow resistance of the bypass gap rose from 18.04 to 26.24 by the optimization. • The bypass ratios at the inlet and outlet were reduced by 36.19% and 14.66%, respectively. -- Abstract: Core bypass flow in block type very high temperature reactor (VHTR) occurs due to the inevitable gaps between the hexagonal core blocks for the block installation and refueling. Since the core bypass flow affects the reactor safety and efficiency, it should be minimized to enhance the core thermal margin. In this regard, the core bypass flow reduction method applying the labyrinth seal mechanism was developed and optimized by using the single-objective shape optimization method. Response surface approximation (RSA) method was adopted as the optimization method. Side wall of the replaceable reflector block was redesigned and response surface approximate model was adopted to optimize the shape of the reflector wall. Computational fluid dynamics (CFD) analyses were carried out not only to assess the limitation of existing method of bypass flow reduction, but also to optimize the design of a newly developed reduction method. The experiment with Seoul National University (SNU) multi-block experimental facility was performed to demonstrate the performance of the reduction method. It was found that the effect of the existing bypass flow reduction method by sealing the bypass gap exit was restricted nearby the lower region of the core. However, the flow resistance factor of the bypass gap increased from 18.04 to 26.24 by the optimized reduction method. The results of the performance test showed that the bypass flow distribution was reduced throughout the entire core regions. The bypass flow ratios at the inlet and the outlet were

  15. A Sub-channel Analysis of a VHTR Fuel Block with Tin Gap-Filler

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Kim, Yong Hee; Yi, Yong Sun; Kim, Hong Pyo

    2005-01-01

    In the Nuclear Hydrogen Development and Demonstration (NHDD) project, two types of VHTRs (Very High Temperature Reactors), prismatic or pebble bed, are under investigation as the nuclear heat source for hydrogen production. In general, the targeted coolant outlet temperature of VHTR is 950∼1000 .deg. C and the maximum allowable fuel temperature is 1250 .deg. C during the normal operation. In the case of the prismatic reactor (PMR), conventional fuel designs result in a small margin in the maximum fuel temperature. This is one of the biggest demerits of the prismatic type In this paper, a technique of lowering the maximum fuel temperature is suggested. The PMR fuel assembly is comprised of many coolant holes and fuel channels. Cylindrical fuel compacts are stacked inside the fuel channel. Consequently, there should be a physical gap between the fuel compact and graphite block, which is filled with the He gas in the conventional design. The heat transfer coefficient of the He gap is very poor, and this increases the fuel temperature substantially. In the proposed design measure, the gap is filled with a liquid metal, tin (Sn) that has a very high thermal conductivity. The effects of tin in the gap with gap distance variation in the viewpoint of thermal hydraulics are quantitatively discussed. Also, the effects of the variations of the axial power distribution are discussed

  16. A dynamic study on the sulfuric acid distillation column for VHTR-assisted hydrogen production systems

    International Nuclear Information System (INIS)

    Youngjoon, Shin; Heesung, Shin; Jiwoon, Jang; Kiyoung, Lee; Jonghwa, Chang

    2007-01-01

    The sulfur-iodine (SI) cycle and the Westinghouse sulfur hybrid cycle coupled to a very high temperature gas-cooled reactor (VHTR) are well known as a feasible technology to produce hydrogen. The concentration of the sulfuric acid solution and its decomposition are essential parts in both cycles. In this paper, the thermophysical properties which are the boiling point, latent heat, and the partial pressures of water, sulfuric acid, and sulfur trioxide have been correlated as a function of the sulfuric acid concentration for the H 2 SO 4 and H 2 O binary chemical system, based on the data in Perry's chemical engineers' hand-book and other experimental data. By using these thermophysical correlations, a dynamic analysis of a sulfuric acid distillation column has been performed to establish the column design requirements and its optimum operation condition. From the results of the dynamic analysis, an optimized column system is anticipated for a distillation column equipped with 2 ideal plates and a second plate feeding system from the bottom plate. The effects of the hold-up of the re-boiler and the reflux ratio from the top product stream on the elapsing time when the system progresses toward a steady state have been analyzed. (authors)

  17. One stacked-column vibration test and analysis for VHTR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Ishizuka, Hiroshi; Ide, Akira; Hayakawa, Hitoshi; Shingai, Kazuteru.

    1978-07-01

    This paper describes experimental results of the vibration test on a single stacked-column and compares them with the analytical results. A 1/2 scale model of the core element of a very high temperature gas-cooled reactor (VHTR) was set on a shaking table. Sinusoidal waves, response time history waves, beat wave and step wave of input acceleration 100 - 900 gal in the frequency of 0.5 to 15 Hz were used to vibrate the table horizontally. Results are as follows: (1) The column has a non-linear resonance and exhibits a hysteresis response with jump points. (2) The column vibration characteristics is similar to that of the finite beams connected with non-linear soft spring. (3) The column resonance frequency decreases with increasing input acceleration. (4) The impact force increases with increasing input acceleration and boundary gap width. (5) Good correlation in vibration behavior of the stacked-column and impact force on the boundary between test and analysis was obtained. (auth.)

  18. PIV Experiments to Measure Flow Phenomena in a Scaled Model of a VHTR Lower Plenum

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy, Jr.; Donald M. McEligot; Richard R. Schultz; Daniel Christensen; Robert J. Pink; Ryan C. Johnson

    2006-09-01

    A report of experimental data collected at the Matched-Index-of-Refraction (MIR) Laboratory in support of contract DE-AC07-05ID14517 and the INL Standard Problem on measurements of flow phenomena occurring in a lower plenum of a typical prismatic VHTR concept reactor to assess CFD code is presented. Background on the experimental setup and procedures is provided along with several samples of data obtained from the 3-D PIV system and an assessment of experimental uncertainty is provided. Data collected in this study include 3-dimensional velocity-field descriptions of the flow in all four inlet jets and the entire lower plenum with inlet jet Reynolds numbers (ReJet) of approximately 4300 and 12,400. These investigations have generated over 2 terabytes of data that has been processed to describe the various velocity components in formats suitable for external release and archived on removable hard disks. The processed data from both experimental studies are available in multi-column text format.

  19. Evaluation of effective coolant flow rate in advanced design of the small scale VHTR core

    International Nuclear Information System (INIS)

    Fumizawa, Motoo; Suzuki, Kunihiko; Murakami, Tomoyuki.

    1988-02-01

    This report describes the evaluation of effective coolant flow rate in the advanced design of the small scale VHTR core. The analytical design study was carried out after the 2nd stage of detailed design in order to reduce the cost of construction. The summary of the analytical results are as follows: (1) Crossflow loss coefficient of flange type fuel block having 0.1 mm of sealing gap is about 100 times higher than that of dowel type block adopted in the 2nd stage of detailed design. (2) In case that coolant channel outer diameter is 52 mm and hydraulic diameter is 6 mm, the effective coolant flow rates using flange and dowel type fuel blocks are 80 % and 70 % respectively. Because the crossflow loss coefficients of dowel type are lower than that of flange type. (3) The effective coolant flow rate, when crossflow loss coefficients are distributed along with the axial direction, agrees well with that using mean value of crossflow loss coefficient i.e. 5 x 10 11 m -4 . (author)

  20. CLUPH: a Fortran program of collision probabilities for hexagonal lattice and its application to VHTR

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Gotoh, Yorio

    1981-02-01

    A new collision probability routine CLUPH was added to the computer program set LAMP-B to analyse the hexagonal VHTR fuel and control blocks where in addition to the annular array of fuel pin rods the asymmetric insertions of burnable poison rods and control rods are characteristic. The perfect reflective boundary condition is no more realistic to consider the arrangement of asymmetric hexagonal blocks. The periodic and the rotational arrangement of blocks are surveyed to consider the interference effect between the burnable poison rods. In addition the effects of coated particle fuel in fuel rod, and of B 4 C grain in burnable poison rod, are investigated. The average cross sections of control rod block were derived from the calculation of a super cell which consists of the control rod block and of the surrounding six fuel blocks. The care was taken to the control rod block located at the core-reflector boundary by replacing a sector of surrounding material in supper cell by reflector material. The two dimensional diffusion calculations of simplified cores of Mk-III were performed to obtain the reactivity worths of control rods, for illustration. (author)

  1. Novel experiments to characterise creep-fatigue degradation in VHTR alloys

    International Nuclear Information System (INIS)

    Simpson, J.A.; Wright, J.K.; Wright, R.N.

    2015-01-01

    It is well known in energy systems that the creep lifetime of high temperature alloys is significantly degraded when a cyclic load is superimposed on components operating in the creep regime. A test method has been developed in an attempt to characterise creep-fatigue behaviour of alloys at high temperature. The test imposes a hold time during the tensile phase of a fully reversed strain-controlled low cycle fatigue test. Stress relaxation occurs during the strain-controlled hold period. This type of fatigue stress relaxation test tends to emphasise the fatigue portion of the total damage and does not necessarily represent the behaviour of a component in-service well. Several different approaches to laboratory testing of creep-fatigue at 950 deg. C have been investigated for Alloy 617, the primary candidate for application in VHTR heat exchangers. The potential for mode switching in a cyclic test from strain control to load control, to allow specimen extension by creep, has been investigated to further emphasise the creep damage. In addition, tests with a lower strain rate during loading have been conducted to examine the influence of creep damage occurring during loading. Very short constant strain hold time tests have also been conducted to examine the influence of the rapid stress relaxation that occurs at the beginning of strain holds. (authors)

  2. Back up core designs for the experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Aochi, Tetsuo; Yasuno, Takehiko; Miyamoto, Yoshiaki; Shindo, Ryuichi; Ikushima, Takeshi

    1979-02-01

    For the Experimental Multi-Purpose Very High Temperature Reactor (thermal power 50 MW and reactor outlet helium temperature 1000 0 C), design studies have been made of two backup cores loaded with new-type fuel elements. The purpose is to improve core operational characteristics, especially in thermohydraulics, of the reference design core consisting of pin-in-block type fuel elements having externally cooled hollow fuel rods. In this report are described the design principles and the analyses made of nuclear, thermal and hydraulic, fuel, and safety performances to determine the backup fuel and core design parameters. The first backup core (SP fuel core) is composed of fuel elements with internally cooled fuel rods (semi-pin), 36 rods in each standard element and 18 rods in each control element. The second backup core (MH fuel core) is composed of multihole fuel elements. 102 fuel and 54 coolant holes in each standard element and 30 fuel and 18 coolant holes in each control element. Either of the cores has 73 fuel columns 4 m high; the arrangement of active core and reactor internal structures is the same as that in the reference design. The backup cores meet nearly all design requirements of the VHTR, permitting the rated power operation with coolant Reynolds number of over 10,000 in the SP core and over 6,000 in the MH core. (author)

  3. IVS Organization

    Science.gov (United States)

    2004-01-01

    International VLBI Service (IVS) is an international collaboration of organizations which operate or support Very Long Baseline Interferometry (VLBI) components. The goals are: To provide a service to support geodetic, geophysical and astrometric research and operational activities. To promote research and development activities in all aspects of the geodetic and astrometric VLBI technique. To interact with the community of users of VLBI products and to integrate VLBI into a global Earth observing system.

  4. Key Factors for the Linkage Strategy between R and D and Commercialization for Gen-ΙV

    International Nuclear Information System (INIS)

    Lee, Kyoungmi; Hong, Jung Suk

    2013-01-01

    The Fukushima nuclear disaster has leaded to enhance the safety and the cost-effectiveness of technology for the future so that advanced countries such as United Sates and France have concerned about a next generation nuclear power plant, Gen-IV(Generation-IV Reactor). Considering various characteristics of nuclear R and D, it is necessary to have more elaborated strategies for the effective development of the next generation of nuclear technology. In this study, we suggest 5 key factors for the successful commercialization of Gen-IV by analyzing the distinct characteristics of nuclear R and D with Gen-IV and CSF(Critical Success Factor)s of several cases in these field and conducting the FGI(Focus Group Interview). Considering these results, we could find and suggest some important points for further strategy for Gen-IV. That is, following five key factors for the linkage improvement between R and D and commercialization of Gen-IV should be considered: the participation of nuclear power plant operators from the beginning, the establishment of consistent and comprehensive plan/roadmap/detailed strategy, the technology development based on global energy issues and international cooperation, the stable and clear funding plans for long-term projects, the cooperation of relative ministries. Gen-IV system is getting a positive response in that it accompanies long-term R and D plans in Korea. We think that the standard of Gen-IV would lead the next generation of nuclear industry if the proper strategy for the cooperation between the private sector and the regulation from the beginning. Moreover, we expect that this study will facilitate its development process from R and D to commercialization

  5. Key Factors for the Linkage Strategy between R and D and Commercialization for Gen-ΙV

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyoungmi; Hong, Jung Suk [Korean Institute of S and T Evaluation and Planning, Seoul (Korea, Republic of)

    2013-05-15

    The Fukushima nuclear disaster has leaded to enhance the safety and the cost-effectiveness of technology for the future so that advanced countries such as United Sates and France have concerned about a next generation nuclear power plant, Gen-IV(Generation-IV Reactor). Considering various characteristics of nuclear R and D, it is necessary to have more elaborated strategies for the effective development of the next generation of nuclear technology. In this study, we suggest 5 key factors for the successful commercialization of Gen-IV by analyzing the distinct characteristics of nuclear R and D with Gen-IV and CSF(Critical Success Factor)s of several cases in these field and conducting the FGI(Focus Group Interview). Considering these results, we could find and suggest some important points for further strategy for Gen-IV. That is, following five key factors for the linkage improvement between R and D and commercialization of Gen-IV should be considered: the participation of nuclear power plant operators from the beginning, the establishment of consistent and comprehensive plan/roadmap/detailed strategy, the technology development based on global energy issues and international cooperation, the stable and clear funding plans for long-term projects, the cooperation of relative ministries. Gen-IV system is getting a positive response in that it accompanies long-term R and D plans in Korea. We think that the standard of Gen-IV would lead the next generation of nuclear industry if the proper strategy for the cooperation between the private sector and the regulation from the beginning. Moreover, we expect that this study will facilitate its development process from R and D to commercialization.

  6. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

    2013-06-01

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  7. Generation-IV nuclear reactors, SFR concept

    International Nuclear Information System (INIS)

    Dufour, P.

    2010-01-01

    In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

  8. CEA and AREVA R and D on V/HTR fuel fabrication with the CAPRI experimental manufacturing line

    International Nuclear Information System (INIS)

    Charollais, Francois; Fonquernie, Sophie; Perrais, Christophe; Perez, Marc; Cellier, Francois; Vitali, Marie-Pierre

    2006-01-01

    In the framework of the French V/HTR fuel development and qualification program, the Commissariat a l'Energie Atomique (CEA) and AREVA through its program called ANTARES (Areva New Technology for Advanced Reactor Energy Supply) conduct R and D projects covering the mastering of UO 2 coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture: - The different stages of UO 2 kernel fabrication GSP Sol-Gel process have been reviewed, reproduced and improved; - The experimental conditions for the chemical vapour deposition (CVD) of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out; - Former CERCA compacting process has been reviewed and updated. In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line). The major objectives of the CAPRI line are: - to recover and validate past knowledge; - to permit the optimisation of reference fabrication processes for kernels and coatings and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating); - to test alternative compact process options; - to fabricate and characterize fuel required for irradiation and qualification purpose; - to specify needs for the fabrication of representative V/HTR TRISO fuel meeting industrial standards. This paper presents the progress status of the R and D conducted on V/HTR fuel particle and compact manufacture by mid 2005. (authors)

  9. Evaluation of the influence of bypass flow gap distribution on the core hot spot in a prismatic VHTR core

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lim, Hong-Sik

    2011-01-01

    Highlights: → A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. → The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. → The predicted gap size is large enough to affect the flow distribution in the core. → The bypass gap and flow distributions are closely related to the local hot spot temperature and its location. → The core restraint mechanism preventing outward movement of graphite block reduces the bypass gap size and hot spot temperature. - Abstract: Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage/swelling and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of fast neutron fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass gap and flow distributions are closely related to the local hot spot and its location and the core restraint mechanism preventing outward movement of the graphite block by a fastening device reduces the bypass gap size, which results in the decrease of maximum fuel temperature not less than 100 deg. C, when compared to the case without it.

  10. Development Of An Experiment For Measuring Flow Phenomena Occurring In A Lower Plenum For VHTR CFD Assessment

    Energy Technology Data Exchange (ETDEWEB)

    D. M. McEligot; K.G. Condie; G. E. Mc Creery; H. M. Mc Ilroy

    2005-09-01

    The objective of the present report is to document the design of our first experiment to measure generic flow phenomena expected to occur in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In the process, fabrication sketches are provided for the use of CFD (computational fluid dynamics) analysts wishing to employ the data for assessment of their proposed codes. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. One aspect of the complex flow in a prismatic VHTR is being addressed: flow and thermal mixing in the lower plenum ("hot streaking" issue). Current prismatic VHTR concepts were examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses were applied to determine key non-dimensional parameters and their magnitudes over this operating range. The flow in the lower plenum can locally be considered to be a situation of multiple jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentum-dominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other.

  11. The shifting study of the active core or a VHTR based on the TRISO packing fraction changing

    Energy Technology Data Exchange (ETDEWEB)

    Silva, F.C.; Pereira, C.; Veloso, M.A.F., E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear. Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Costa, A.L. [Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (CNPq), Brasilia, DF (Brazil)

    2011-07-01

    A simplified VHTR core was analyzed, loaded with a fuel mixture of uranium oxide together with reprocessed transuranic nuclides. The TRUs were reprocessed together with Pu, Am, Np and Cm (23.80%) from PWR spent fuel, and dissolved in depleted uranium (0.2% {sup 235}U) until obtain 15% LEU-fuel ({sup 235}U, {sup 239}Pu and {sup 241}Pu). The shifting study of the active core was based on changes in the TRISO particle. Five cases were analyzed changing the VM/VF ratio (moderator volume/ fuel volume) making changes in the TRISO packing fraction (tpf), where tpf represents the ratio of TRISO particle on the fuel pin. The fuels were evaluated during the burnup up to 100,000.0 MWd/THM, during 990 days and without reloads. Then, it evaluated the multiplication (k{sub eff}) at zero and full power, fuel temperature coefficient ({alpha}{sub TF}), moderator temperature coefficient ({alpha}{sub TM}), and fuel composition at BOL (begin of life) and EOL (end of life), using the code Winfrith Improved Multi-Group Scheme (WIMSD5). The results show an overall heavy metal decrease in relation to the total TRU, with some Pu and Np being transmuted in the VHTR core. The results also clearly show the advantage of using reprocessed fuel in VHTR. It decreases the impact of the final spent fuel deposition, minimizes the cost of new fuel using reprocessed fuel and depleted uranium and demonstrated the promising neutronic behavior of the new types of nuclear reactors. (author)

  12. Experimental Investigation on Cross Flow of Wedge-shaped Gap in the core of Prismatic VHTR

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Park, Goon Cherl; Cho, Hyoung Kyu; Yoon, Su Jong

    2014-01-01

    The core of the PMR type reactor consists of assemblies of hexagonal graphite blocks. The graphite blocks have lots of advantages for neutron economy and high temperature structural integrity. The height and flat-to-flat width of fuel bock are 793 mm and 360 mm, respectively. Each block has 108 coolant channels of which the diameter is 16 mm. And there are gaps between blocks not only vertically but also horizontally for reloading of the fuel elements. The vertical gap induces the bypass flow and through the horizontal gap the cross flow is formed. Since the complicated flow distribution occurs by the bypass flow and cross flow, flow characteristics in the core of the PMR reactor cannot be treated as a simple pipe flow. The fuel zone of the PMR core consists of multiple layers of fuel blocks. The shape change of the fuel blocks could be caused by the thermal expansion and fast-neutron induced shrinkage. It could make different axial shrinkage of fuel block and this leads to wedge-shaped gaps between two stacked fuel blocks. The cross flow is often considered as a leakage flow through the horizontal gap between stacked fuel blocks and it complicates the flow distribution in the reactor core by connecting the coolant channel and the bypass gap. Moreover, the cross flow could lead to uneven coolant distribution and consequently cause superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. To develop the cross flow loss coefficient model for determination of the flow distribution for PMR core analysis codes, study on cross flow for PMR200 core is essential. In particular, to predict the amount of flow through the cross flow gap, obtaining accurate flow loss coefficient is important. In this study, the full-scale cross flow experimental facility was constructed to

  13. A Design of He-Molten Salt Intermediate Heat Exchanger for VHTR

    International Nuclear Information System (INIS)

    Jeong, Hui Seong; Bang, Kwang Hyun

    2010-01-01

    The Very High Temperature Reactor (VHTR), one of the most challenging next generation nuclear reactors, has recently drawn an international interest due to its higher efficiency and the operating conditions adequate for supplying process heat to the hydrogen production facilities. To make the design of VHTR complete and plausible, the designs of the Intermediate Heat Transport Loop (IHTL) as well as the Intermediate Heat Exchanger (IHX) are known to be one of the difficult engineering tasks due to its high temperature operating condition (up to 950 .deg. C). A type of compact heat exchangers such as printed circuit heat exchanger (PCHE) has been recommended for the IHX in the technical and economical respects. Selection of the heat transporting fluid for the intermediate heat transport loop is important in consideration of safety and economical aspects. Although helium is currently the primary interest for the intermediate loop fluid, several safety concerns of gas fluids have been expressed. If the pressure boundary of the intermediate loop fails, the blowdown of the gas may overcool the reactor core and then the heat sink is lost after the blowdown. Also the large inventory of gas in the intermediate loop may leak into the primary side. There is also a recommendation that the nuclear plant and the hydrogen production plant be separated by a certain distance to ensure the safety of the nuclear plant in case of accidental heavy gas release from the chemical plant. In this circumstance, the pumping power of gas fluid in the intermediate loop will be large enough to degrade the economics of nuclear hydrogen.An alternative candidate for the intermediate loop fluid in consideration of these safety and economical problems of gas fluid can be molten salts. The Flinak molten salt, a eutectic mixture of LiF, NaF and KF (46.5:11.5:42.0 mole %) is considered to be a potential candidate for the heat transporting fluid in the IHTL due to its chemical stability against the

  14. Reducing uncertainty in personnel dosimetry calculations in the VHTR plant using MAVRIC

    International Nuclear Information System (INIS)

    Flaspoehler, T.; Petrovic, B.

    2013-01-01

    This work analyzes the efficacy of the MAVRIC sequence of the Scale 6.1 code package with respect to the accuracy of results and the ability to utilize large-memory, parallel machines. MAVRIC implements the hybrid FW-CADIS methodology to solve neutron and photon transport for shielding applications. Using the discrete ordinates method to solve the Boltzmann transport equation, an importance map is generated which MAVRIC then uses to bias a stochastic Monte Carlo simulation. The MAVRIC sequence is applied to generate neutron and photon dose rate distributions of improved accuracy in a model of a proposed VHTR power plant. Problems like this one, with a size on the order of magnitude of a nuclear power plant, require a prohibitive amount of memory to store complete importance maps. The issue is addressed by refining the mesh in areas around the source through the detector regions, while leaving a coarse mesh elsewhere. Additionally through the use of parallel computing, the angular flux can be expanded in higher quadrature sets, which leads to a better importance map while requiring no extra memory requirements during the Monte Carlo portion of the sequence. The final Monte Carlo simulations can be run concurrently on several machines with results combined after the fact, emulating parallelism that is not yet available in MAVRIC sequence. Using a combination of strategies, the MAVRIC sequence is shown to be able to scale across available computational resources, allowing the user to more quickly obtain Monte Carlo results with lower relative uncertainties in large, deep-penetration shielding problems. (authors)

  15. CFD analysis of the VHTR prismatic core with variation of geometry parameters

    Energy Technology Data Exchange (ETDEWEB)

    Lira, Carlos A.B.O.; Paiva, Pedro P.D.S., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

    2017-11-01

    The Very High Temperature Reactor is a thermal, graphite moderated and helium cooled nuclear reactor. The purpose of this work is to study the behavior of the VHTR by means of parametric analysis, altering the energy generation profile in the fuel blocks and the influence of modifications in the geometry itself. The coolant flow through the coolant channels and by-pass channels were analyzed in a 1/12{sup th} section of a fuel block column. Geometry was used with by-pass channels of different dimensions, besides one that had only the cooling channels, without by-pass channel. It has been found that the existence of a by-pass flow induces an increase in the temperature gradient in the fuel block. Comparative studies were performed between the results obtained in simulations carried out with different profiles of thermal energy generation (uniform and sinusoidal) in the fuel channels. It was verified that when there is the same total thermal energy generation in the fuel block, the maximum temperature observed in each of the materials is smaller for the generation with sinusoidal profile. Computer simulations were performed using a geometry with a central channel with the same diameter as the others to verify the hypothesis that the existence of a temperature gradient in the fuel block, with the highest temperature at the center and the lowest temperature being at the periphery of this block, is due to the smaller dimension of the coolant channel located in the center of this block. The results obtained confirm the hypothesis. (author)

  16. Numerical solution of heat transfer process in a prismatic VHTR core accompanying bypass and cross flows

    International Nuclear Information System (INIS)

    Wang, Li; Liu, Qiusheng; Fukuda, Katsuya

    2016-01-01

    Highlights: • Three-dimensional CFD analysis is conducted for the thermal analysis in the reactor core. • Hot spot temperature, coolant channel outlet temperature distribution are affected by bypass flow. • Bypass gap size has significant influence on temperature and flow distribution in the core. • Cross flow has some effect on the temperature distribution of the coolant in the core due to flow mixing in the cross gaps. - Abstract: Bypass flow and cross flow gaps both exist in the core of a very high temperature gas-cooled reactor (VHTR), which is inevitable owing to tolerances in manufacturing, thermal expansion and irradiation shrinkage. The coolant mass flow rate distribution, temperature distribution, and hot spot temperature are significantly affected by bypass and cross flows. In the present study, three-dimensional CFD analysis is conducted for thermal analysis of the reactor core. A validation study for the turbulence model is performed by comparing the friction coefficient with published correlations. A sensitivity study of the near wall mesh is conducted to ensure mesh quality. Parametric studies are performed by changing the size of the bypass and cross gaps using a one-twelfth sector of a fuel block. Simulation results show the influence of the bypass gap size on temperature distribution and coolant mass flow rate distribution in the prismatic core. It is shown that the maximum fuel and coolant channel outlet temperatures increase with an increase in the gap size, which may lead to a structural risk to the fuel block. The cross flow is divided into two types: the cross flow from the bypass gap to the coolant channels and the cross flow from the high-pressure coolant channels to low-pressure coolant channels. These two types of flow have an opposing influence on the temperature gradient. It is found that the presence of the cross flow gaps may have a significant effect on the distribution of the coolant in the core due to flow mixing in the

  17. Preliminary Overview of a Helium Cooling System for the Secondary Helium Loop in VHTR-based SI Hydrogen Production Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Cho, Mintaek; Kim, Dahee; Lee, Taehoon; Lee, Kiyoung; Kim, Yongwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Nuclear hydrogen production facilities consist of a very high temperature gas-cooled nuclear reactor (VHTR) system, intermediate heat exchanger (IHX) system, and a sulfur-iodine (SI) thermochemical process. This study focuses on the coupling system between the IHX system and SI thermochemical process. To prevent the propagation of the thermal disturbance owing to the abnormal operation of the SI process components from the IHX system to the VHTR system, a helium cooling system for the secondary helium of the IHX is required. In this paper, the helium cooling system has been studied. The temperature fluctuation of the secondary helium owing to the abnormal operation of the SI process was then calculated based on the proposed coupling system model. Finally, the preliminary conceptual design of the helium cooling system with a steam generator and forced-draft air-cooled heat exchanger to mitigate the thermal disturbance has been carried out. A conceptual flow diagram of a helium cooling system between the IHX and SI thermochemical processes in VHTR-based SI hydrogen production facilities has been proposed. A helium cooling system for the secondary helium of the IHX in this flow diagram prevents the propagation of the thermal disturbance from the IHX system to the VHTR system, owing to the abnormal operation of the SI process components. As a result of a dynamic simulation to anticipate the fluctuations of the secondary helium temperature owing to the abnormal operation of the SI process components with a hydrogen production rate of 60 mol·H{sub 2}/s, it is recommended that the maximum helium cooling capacity to recover the normal operation temperature of 450 .deg. C is 31,933.4 kJ/s. To satisfy this helium cooling capacity, a U-type steam generator, which has a heat transfer area of 12 m{sup 2}, and a forced-draft air-cooled condenser, which has a heat transfer area of 12,388.67 m{sup 2}, are required for the secondary helium cooling system.

  18. Nordic forum for generation IV reactors, status and activities in 2012

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.; Lauritzen, B.; Nonboel, E.

    2012-12-01

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

  19. Nordic forum for generation IV reactors, status and activities in 2012

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. [Institutt for Energiteknikk, OECD Halden Reactor Project, Kjeller (Norway); Lauritzen, B.; Nonboel, E. [Technical Univ. of Denmark. DTU Nutech, Roskilde (Denmark)

    2012-12-15

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

  20. Corrosion of nickel-base heat resistant alloys in simulated VHTR coolant helium at very high temperatures

    International Nuclear Information System (INIS)

    Shindo, Masami; Kondo, Tatsuo

    1976-01-01

    A comparative evaluation was made on three commercial nickel-base heat resistant alloys exposed to helium-base atmosphere at 1000 0 C, which contained several impurities in simulating the helium cooled very high temperature nuclear reactor (VHTR) environment. The choice of alloys was made so that the effect of elements commonly found in commercial alloys were typically examined. The corrosion in helium at 1000 0 C was characterized by the sharp selection of thermodynamically unstable elements in the oxidizing process and the resultant intergranular penetration and internal oxidation. Ni-Cr-Mo-W type solution hardened alloy such as Hastelloy-X showed comparatively good resistance. The alloy containing Al and Ti such as Inconel-617 suffered adverse effect in contrast to its good resistance to air oxidation. The alloy nominally composed only of noble elements, Ni, Fe and Mo, such as Hastelloy-B showed least apparent corrosion, while suffered internal oxidation due to small amount of active impurities commonly existing in commercial heats. The results were discussed in terms of selection and improvement of alloys for uses in VHTR and the similar systems. (auth.)

  1. Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

  2. Experimental study of core bypass flow in a prismatic VHTR based on a two-layer block model

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Huhu, E-mail: huhuwang@tamu.edu; Hassan, Yassin A., E-mail: y-hassan@tamu.edu; Dominguez-Ontiveros, Elvis, E-mail: elvisdom@tamu.edu

    2016-09-15

    Bypass flow in a prismatic very high temperature gas-cooled nuclear reactor (VHTR) plays an important role in determining the coolant distribution in the core region. Efficient removal of heat from the core relies on the majority of coolant passing through the coolant channels instead of the bypass gaps. Consequently, the bypass flow fraction and its flow characteristic are important in the design process of the prismatic VHTR. The objective of this study is to experimentally investigate the flow behavior including the turbulence characteristics inside the bypass gaps using laser Doppler velocimetry (LDV), bypass fraction and pressure drops in the system. The experiment facility constructed at Texas A&M University is a scaled model consisting of two layers of fuel blocks. The distributions of the mean streamwise velocity, turbulence intensity and turbulence kinetic energy within the bypass gap at two different elevations under different Reynolds number were investigated. Uncertainties in the bypass flow fraction estimation were evaluated. The velocity and turbulence study in this work is considered to be unique, and may serve as a benchmark for the related numerical calculations.

  3. Overview of materials R and D for fusion and Gen-4

    Energy Technology Data Exchange (ETDEWEB)

    Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan); Tavassoli, F.; Carre, F.; Billot, P. [CEA Saclay, 91 - Gif sur Yvette (France); Zinide, S. [Oak Ridge National Laboratory, Materials Science and Technology Div., AK TN (United States)

    2007-07-01

    Full text of publication follows: In view of the growing need for energy, the risk of exhaustion of fossil fuel and the problem of global warming, the nuclear energy is receiving added attention as a realistic and viable advanced solution. International collaborations on Generation IV (Gen-IV) fission reactors and on ITER and DEMO fusion reactors are developing. This is particularly the case in the sector of materials, where they hold the key to success of these systems. The international community has recognized and planned its materials R and D work for Fusion and Gen-IV reactors with the following considerations: 1- The time allotted to materials R and D is short and may not allow development of totally new materials. 2- Activities required, to cover existing materials variations and service conditions necessary for reactor design, are very time consuming. 3- The work to be done must build upon the existing knowledge of materials and avoid duplications. Although ITER for fusion and Generation four International Forum (GIF) for Gen-IV are important international collaborative programs, they are insufficient to meet all the national energy policies of the participating countries. This paper provides an overview of the materials R and D carried out for fusion and Gen-IV reactors at international and national levels. Materials programs discussed include both cross-cutting and reactor specific actions, where major tasks can be defined as: + Cross-cutting materials tasks: - materials for high temperature service; - materials with neutron damage tolerance; - materials behavior analysis and modeling; - high temperature design methodology. + Reactor specific materials tasks: - very high temperature alloys; - carbon, high temperature ceramics and their composites; - materials compatibilities. Starting with a brief introduction of materials R and D strategies, ITER and Broader Approach (BA), overall activities for fusion and GIF for Gen-IV will be reviewed. Domestic

  4. Asteroids IV

    Science.gov (United States)

    Michel, Patrick; DeMeo, Francesca E.; Bottke, William F.

    . Asteroids, like planets, are driven by a great variety of both dynamical and physical mechanisms. In fact, images sent back by space missions show a collection of small worlds whose characteristics seem designed to overthrow our preconceived notions. Given their wide range of sizes and surface compositions, it is clear that many formed in very different places and at different times within the solar nebula. These characteristics make them an exciting challenge for researchers who crave complex problems. The return of samples from these bodies may ultimately be needed to provide us with solutions. In the book Asteroids IV, the editors and authors have taken major strides in the long journey toward a much deeper understanding of our fascinating planetary ancestors. This book reviews major advances in 43 chapters that have been written and reviewed by a team of more than 200 international authorities in asteroids. It is aimed to be as comprehensive as possible while also remaining accessible to students and researchers who are interested in learning about these small but nonetheless important worlds. We hope this volume will serve as a leading reference on the topic of asteroids for the decade to come. We are deeply indebted to the many authors and referees for their tremendous efforts in helping us create Asteroids IV. We also thank the members of the Asteroids IV scientific organizing committee for helping us shape the structure and content of the book. The conference associated with the book, "Asteroids Comets Meteors 2014" held June 30-July 4, 2014, in Helsinki, Finland, did an outstanding job of demonstrating how much progress we have made in the field over the last decade. We are extremely grateful to our host Karri Muinonnen and his team. The editors are also grateful to the Asteroids IV production staff, namely Renée Dotson and her colleagues at the Lunar and Planetary Institute, for their efforts, their invaluable assistance, and their enthusiasm; they made life as

  5. Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-09-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

  6. Advanced construction materials for thermo-chemical hydrogen production from VHTR process heat

    International Nuclear Information System (INIS)

    Kosmidou, Theodora; Haehner, Peter

    2009-01-01

    The (very) high temperature reactor concept ((V)HTR) is characterized by its potential for process heat applications. The production of hydrogen by means of thermo-chemical cycles is an appealing example, since it is more efficient than electrolysis due to the direct use of process heat. The sulfur-iodine cycle is one of the best studied processes for the production of hydrogen, and solar or nuclear energy can be used as a heating source for the high temperature reaction of this process. The chemical reactions involved in the cycle are: I 2 (l) + SO 2 (g) +2 H 2 O (l) → 2HI (l) + H 2 SO 4 (l) (70-120 deg. C); H 2 SO 4 (l) → H 2 O (l) + SO 2 (g) + 1/2 O 2 (g) (800-900 deg. C); 2HI (l) → I 2 (g) + H 2 (g) (300-450 deg. C) The high temperature decomposition of sulphuric acid, which is the most endothermic reaction, results in a very aggressive chemical environment which is why suitable materials for the decomposer heat exchanger have to be identified. The class of candidate materials for the decomposer is based on SiC. In the current study, SiC based materials were tested in order to determine the residual mechanical properties (flexural strength and bending modulus, interfacial strength of brazed joints), after exposure to an SO 2 rich environment, simulating the conditions in the hydrogen production plant. Brazed SiC specimens were tested after 20, 100, 500 and 1000 hrs exposure to SO 2 rich environment at 850 o C under atmospheric pressure. The gas composition in the corrosion rig was: 9.9 H 2 O, 12.25 SO 2 , 6.13 O 2 , balance N 2 (% mol). The characterization involved: weight change monitoring, SEM microstructural analysis and four-point bending tests after exposure. Most of the specimens gained weight due to the formation of a corrosion layer as observed in the SEM. The corrosion treatment also showed an effect on the mechanical properties. In the four-point bending tests performed at room temperature and at 850 deg. C, a decrease in bending modulus with

  7. FutureGen Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Cabe, Jim; Elliott, Mike

    2010-09-30

    This report summarizes the comprehensive siting, permitting, engineering, design, and costing activities completed by the FutureGen Industrial Alliance, the Department of Energy, and associated supporting subcontractors to develop a first of a kind near zero emissions integrated gasification combined cycle power plant and carbon capture and storage project (IGCC-CCS). With the goal to design, build, and reliably operate the first IGCC-CCS facility, FutureGen would have been the lowest emitting pulverized coal power plant in the world, while providing a timely and relevant basis for coal combustion power plants deploying carbon capture in the future. The content of this report summarizes key findings and results of applicable project evaluations; modeling, design, and engineering assessments; cost estimate reports; and schedule and risk mitigation from initiation of the FutureGen project through final flow sheet analyses including capital and operating reports completed under DOE award DE-FE0000587. This project report necessarily builds upon previously completed siting, design, and development work executed under DOE award DE-FC26- 06NT4207 which included the siting process; environmental permitting, compliance, and mitigation under the National Environmental Policy Act; and development of conceptual and design basis documentation for the FutureGen plant. For completeness, the report includes as attachments the siting and design basis documents, as well as the source documentation for the following: • Site evaluation and selection process and environmental characterization • Underground Injection Control (UIC) Permit Application including well design and subsurface modeling • FutureGen IGCC-CCS Design Basis Document • Process evaluations and technology selection via Illinois Clean Coal Review Board Technical Report • Process flow diagrams and heat/material balance for slurry-fed gasifier configuration • Process flow diagrams and heat/material balance

  8. Influence of heating rate on corrosion behavior of Ni-base heat resistant alloys in simulated VHTR helium environment

    International Nuclear Information System (INIS)

    Kurata, Yuji; Kondo, Tatsuo

    1985-04-01

    The influence of heating rate on corrosion and carbon transfer was studied for Ni-base heat resistant alloys exposed to simulated VHTR(very high temperature reactor) coolant environment. Special attention was focused to relationship between oxidation and carburization at early stage of exposure. Tests were conducted on two heats of Hastelloy XR with different boron(B) content and the developmental alloys, 113MA and KSN. Two kinds of heating rates, i.e. 80 0 C/min and 2 0 C/min, were employed. Corrosion tests were carried out at 900 0 C up to 500 h in JAERI Type B helium, one of the simulated VHTR primary coolant specifications. Under higher heating rate, oxidation resistance of both heats of Hastelloy XR(2.8 ppmB and 40 ppmB) were equivalent and among the best, then KSN and 113MA followed in the order. Under lower heating rate only alloy, i.e. Hastelloy XR with 2.8 ppmB, showed some deteriorated oxidation resistance while all others being unaffected by the heating rate. On the other hand the carbon transfer behavior showed strong dependence on the heating rate. In case of higher heating rate, significant carburization occured at early stage of exposure and thereafter the progress of carburization was slow in all the alloys. On the other hand only slow carburization was the case throughout the exposure in case of lower heating rate. The carburization in VHTR helium environment was interpreted as to be affected by oxide film formation in the early stage of exposure. The carbon pick-up was largest in Hastelloy XR with 40 ppmB and it was followed by Hastelloy XR with 2.8 ppmB. 113MA and KSN were carburized only slightly. The observed difference of carbon pick-up among the alloys tested was interpreted to be attributed mainly to the difference of the carbon activity, the carbide precipitation characteristics among the alloys tested. (author)

  9. Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

    Science.gov (United States)

    Valentin Rodriguez, Francisco Ivan

    High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat

  10. Algoritmos genéticos locales

    OpenAIRE

    García-Martínez, Carlos; Lozano, Manuel

    2007-01-01

    Los Algoritmos Genéticos Locales son procedimientos que iterativamente re nan soluciones dadas. Su diferencia con procedimientos de mejora iterativa clásicos reside en el uso de operadores genéticos para realizar el re namiento. En este estudio presentamos un nuevo Algoritmo Genético Local Binario basado en un Algoritmo Genético Estacionario. Hemos comparado el Algoritmo Genético Local Binario con otros procedimientos de mejora iterativa de la literatura. Los res...

  11. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  12. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    International Nuclear Information System (INIS)

    Kastenberg, William E.; Blandford, Edward; Kim, Lance

    2009-01-01

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public

  13. High Temperature Degradation Behavior and its Mechanical Properties of Inconel 617 alloy for Intermediate Heat Exchanger of VHTR

    International Nuclear Information System (INIS)

    Jo, Tae Sun; Kim, Se Hoon; Kim, Young Do; Park, Ji Yeon

    2008-01-01

    Inconel 617 alloy is a candidate material of intermediate heat exchanger (IHX) and hot gas duct (HGD) for very high temperature reactor (VHTR) because of its excellent strength, creep-rupture strength, stability and oxidation resistance at high temperature. Among the alloying elements in Inconel 617, chromium (Cr) and aluminum (Al) can form dense oxide that act as a protective surface layer against degradation. This alloy supports severe operating conditions of pressure over 8 MPa and 950 .deg. C in He gas with some impurities. Thus, high temperature stability of Inconel 617 is very important. In this work, the oxidation behavior of Inconel 617 alloy was studied by exposure at high temperature and was discussed the high temperature degradation behavior with microstructural changes during the surface oxidation

  14. GenLab, Laboratorio Virtual de Genética

    Directory of Open Access Journals (Sweden)

    Fidel Ramírez

    2000-07-01

    Full Text Available GenLab es el nombre que tiene el software diseñado por nosotros, en el cual se modela el proceso meiótico y la fecundación en organismos diploides. El objetivo de esta aplicación es ilustrar el resultado de un cruce determinado, tratando de ser lo más ajustados a la realidad. La modelación de la reproducción sexual se realiza internamente y el GenLab se limita a presentar los resultados según el número de descendencia seleccionado para un cruce específico, esto significa que se puede escoger una gran cantidad de características para los parentales y se puede estudiar la frecuencia de estos en la descendencia. El modelo cuenta con base de datos donde están almacenados algunos de los locus de Drosophila melanogaster junto con su ubicación en centimorgans 1. EI propósito de este modelo es servir como herramienta pedagógica  y didáctica tanto en universidades como en colegios, facilitando el aprendizaje de algunos principios básicos de la genética, por lo cual puede ser usado si se cuenta con una conexión a Internet y un navegador visitando http://biologia.unal.edu.co/fidel.

  15. Taxonomic dissection of the genus Micrococcus: Kocuria gen. nov., Nesterenkonia gen. nov., Kytococcus gen. nov., Dermacoccus gen. nov., and Micrococcus Cohn 1872 gen. emend.

    Science.gov (United States)

    Stackebrandt, E; Koch, C; Gvozdiak, O; Schumann, P

    1995-10-01

    The results of a phylogenetic and chemotaxonomic analysis of the genus Micrococcus indicated that it is significantly heterogeneous. Except for Micrococcus lylae, no species groups phylogenetically with the type species of the genus, Micrococcus luteus. The other members of the genus form three separate phylogenetic lines which on the basis of chemotaxonomic properties can be assigned to four genera. These genera are the genus Kocuria gen. nov. for Micrococcus roseus, Micrococcus varians, and Micrococcus kristinae, described as Kocuria rosea comb. nov., Kocuria varians comb. nov., and Kocuria kristinae comb. nov., respectively; the genus Nesterenkonia gen. nov. for Micrococcus halobius, described as Nesterenkonia halobia comb. nov.; the genus Nesterenkonia gen. nov. for Micrococcus halobius, described as Nesterenkonia halobia comb. nov.; the genus Dermacoccus gen. nov. for Micrococcus nishinomiyaensis, described as Dermacoccus nishinomiyaensis comb. nov.; and the genus Kytocossus gen. nov. for Micrococcus sedentarius, described as Kytococcus sedentarius comb. nov. M. luteus and M. lylae, which are closely related phylogenetically but differ in some chemotaxonomic properties, are the only species that remain in the genus Micrococcus Cohn 1872. An emended description of the genus Micrococcus is given [corrected].

  16. Pebble bed modular reactor - The first Generation IV reactor to be constructed

    International Nuclear Information System (INIS)

    Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

    2004-01-01

    Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

  17. Nordic Nuclear Materials Forum for Generation IV Reactors

    International Nuclear Information System (INIS)

    Anghel, C.; Penttilae, S.

    2010-03-01

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  18. Nordic Nuclear Materials Forum for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anghel, C. (Studsvik Nuclear AB, Nykoeping (Sweden)); Penttilae, S. (Technical Research Centre of Finland, VTT (Finland))

    2010-03-15

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  19. Unsteady Reynolds Averaged Navier-Stokes and Large Eddy Simulations of Flows across Staggered Tube Bundle for a VHTR Lower Plenum Design

    International Nuclear Information System (INIS)

    Choi, Hyeon Kyeong; Park, Jong Woon

    2013-01-01

    In this work, behavior of unsteady and oscillating flow through a typical tube bundle array are analyzed by unsteady computations: 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) and the results are compared with existing experimental data. In order to confirm appropriateness and limitations of CFD applications in the Korean VHTR design, two types of unsteady computations are performed such as 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) for the existing tube bundle array. The velocity component profiles are compared with the experimental data and it is concluded that the URANS with the standard k-ω model is reasonably appropriate for cost-effective VHTR lower plenum analysis. Nevertheless, if more accurate results are needed, the LES-Smagorinsky computation is recommended considering limitations in the time averaged RANS in capturing small eddies

  20. From AWE-GEN to AWE-GEN-2d: a high spatial and temporal resolution weather generator

    Science.gov (United States)

    Peleg, Nadav; Fatichi, Simone; Paschalis, Athanasios; Molnar, Peter; Burlando, Paolo

    2016-04-01

    A new weather generator, AWE-GEN-2d (Advanced WEather GENerator for 2-Dimension grid) is developed following the philosophy of combining physical and stochastic approaches to simulate meteorological variables at high spatial and temporal resolution (e.g. 2 km x 2 km and 5 min for precipitation and cloud cover and 100 m x 100 m and 1 h for other variables variable (temperature, solar radiation, vapor pressure, atmospheric pressure and near-surface wind). The model is suitable to investigate the impacts of climate variability, temporal and spatial resolutions of forcing on hydrological, ecological, agricultural and geomorphological impacts studies. Using appropriate parameterization the model can be used in the context of climate change. Here we present the model technical structure of AWE-GEN-2d, which is a substantial evolution of four preceding models (i) the hourly-point scale Advanced WEather GENerator (AWE-GEN) presented by Fatichi et al. (2011, Adv. Water Resour.) (ii) the Space-Time Realizations of Areal Precipitation (STREAP) model introduced by Paschalis et al. (2013, Water Resour. Res.), (iii) the High-Resolution Synoptically conditioned Weather Generator developed by Peleg and Morin (2014, Water Resour. Res.), and (iv) the Wind-field Interpolation by Non Divergent Schemes presented by Burlando et al. (2007, Boundary-Layer Meteorol.). The AWE-GEN-2d is relatively parsimonious in terms of computational demand and allows generating many stochastic realizations of current and projected climates in an efficient way. An example of model application and testing is presented with reference to a case study in the Wallis region, a complex orography terrain in the Swiss Alps.

  1. The generation IV nuclear reactor systems - Energy of future

    International Nuclear Information System (INIS)

    Ohai, Dumitru; Jianu, Adrian

    2006-01-01

    Ten nations joined within the Generation IV International Forum (GIF), agreeing on a framework for international cooperation in research. Their goal is to develop future-generation nuclear energy systems that can be licensed, constructed, and operated in an economically competitive way while addressing the issues of safety, proliferation, and other public perception concerns. The objective is for the Gen IV systems to be available for deployment by 2030. Using more than 100 nuclear experts from its 10 member nations, the GIF has developed a Gen IV Technology Roadmap to guide the research and development of the world's most advanced, efficient and safe nuclear power systems. The Gen IV Technology Roadmap calls for extensive research and development of six different potential future reactor systems. These include water-cooled, gas-cooled, liquid metal-cooled and nonclassical systems. One or more of these reactor systems will provide the best combination of safety, reliability, efficiency and proliferation resistance at a competitive cost. The main goals for the Gen IV Nuclear Energy Systems are: - Provide sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel use for worldwide energy production; - Minimize and manage their nuclear waste and noticeably reduce the long-term stewardship burden in the future, improving the protection of public health and the environment; - Increase the assurance that these reactors are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased protection against acts of terrorism; - Have a clear life-cycle cost advantage over other energy sources; - Have a level of financial risk comparable to other energy projects; - Excel in safety and reliability; - Have a low likelihood and degree of reactor core damage. (authors)

  2. Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

    International Nuclear Information System (INIS)

    Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun; Sun, Xiaodong; Christensen, Richard N.; Oh, Chang H.

    2015-01-01

    Highlights: • A 1/8th geometric-scale test facility that models the VHTR hot plenum is proposed. • Geometric scaling analysis is introduced for VHTR to analyze air-ingress accident. • Design calculations are performed to show that accident phenomenology is preserved. • Some analyses include time scale, hydraulic similarity and power scaling analysis. • Test facility has been constructed and shake-down tests are currently being carried out. - Abstract: A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to depend largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time

  3. Algoritmos genéticos

    Directory of Open Access Journals (Sweden)

    José Jesús Martínez Páez

    1998-10-01

    Full Text Available Esta técnica se basa en el concepto de evolución a través de selección de los mejores individuos, y de los operadores genéticos de selección, reproducción y mutación. Se trata entonces, de definir un espacio de soluciones para el problema que se quiere solucionar, en una cadena de bits. A esto se le conoce como la codificación del cromosoma, donde cada bit, denominado gen  tiene cierto significado especial. Inicialmente el algoritmo genera al azar muchas de estas cadenas o seres, es decir, una población, que luego confronta can un ambiente, que es el problema solucionar o función que se quiere optimizar. De esta confrontación  o evaluación a que se somete cada ser. Se obtiene información sobre cómo se comporto cada uno. A través de métodos aleatorios, pero con probabilidad de selección proporcional a su comportamiento, es decir, a mejor comportamiento mayor probabilidad, se selecciona una nueva población de seres supuestamente mejores que la generación anterior.

  4. Cyclic Deformation and Fatigue Behaviors of Alloy 617 Base Metal and Weldments at 900℃ for VHTR Applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seon Jin; Kim, Byung Tak; Dewa, Rando T.; Hwang, Jeong Jun; Kim, Tae Su [Pukyong National Univ., Busan (Korea, Republic of); Kim, Woo Gon; Kim, Eung Seon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    An analysis of cyclic deformation can contribute to a deeper understanding of the fatigue fracture mechanisms as well as to improvements in the design and application of VHTR system. However, the studies associated with cyclic deformation and low cycle fatigue (LCF) properties of Alloy 617 have focused mainly on the base metal, with little attention given to the weldments. Totemeier studied on high-temperature creep-fatigue of Alloy 617 base metal and weldments. Current research activities at PKNU and KAERI focus on the study of cyclic deformation and LCF behaviors of Alloy 617 base metal (BM) and weldments (WM) specimens were machined from GTAW buttwelded plates at very high-temperature of 900℃. In this work, the cyclic deformation characteristics and fatigue behaviors of Alloy 617 BM and WM are studied and discussed with respect to LCF. In this paper, cyclic deformation and low cycle fatigue behaviors of Alloy 617 base metal and weldments was evaluated using strain-controlled LCF tests at 900℃for 0.6% total strain range. Results of the current experiments can be concluded; The WM specimen has shown a higher cyclic stress response than the BM specimen. The fatigue life of WM specimen was reduced relative to that of BM specimen.

  5. A Dynamic Simulation Program for a Hydriodic Acid Concentration and Decomposition Process in the VHTR-SI Process

    International Nuclear Information System (INIS)

    Chang, Ji Woon; Shin, Young Joon; Lee, Tae Hoon; Lee, Ki Young; Kim, Yong Wan; Chang, Jong Hwa; Youn, Cheung

    2011-01-01

    The Sulfur-Iodine (SI) cycle which can produce hydrogen by using nuclear heat consists of a Bunsen reaction (Section 1), a sulfur acid concentration and decomposition (Section 2), and a hydriodic acid concentration and decomposition (Section 3). The heat required in the SI process can be supplied through an intermediate heat exchanger (IHX) by a Very High Temperature Gas Cooled Reactor (VHTR). The Korea Atomic Energy Research Institute-Dynamic Simulation Code (KAERI-DySCo) based on the Visual C++ is an integration application software that simulates the dynamic behavior of the SI process. KAERI-DySCo was prepared to solve dynamic problem of the seven chemical reactors which consist of Sections 2 and 3. Section 3 is the key part of the SI process, because the strong non-ideality and the partial immiscibility of the binary HI.H 2 O and the ternary HI.I 2 .H 2 O (HIX solution) mixture make it difficult to model and simulate the dynamic behavior of the system. Therefore, it is necessary to compose separately a dynamic simulation program for Section 3 in KAERI-DySCo optimization. In this paper, a simulation program to analyze the dynamic behavior of Section 3 is introduced using the prepared KAERI-DySCo, and results of dynamic simulation are represented by running the program

  6. CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

    2010-01-01

    The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

  7. IV treatment at home

    Science.gov (United States)

    ... Other IV treatments you may receive after you leave the hospital include: Treatment for hormone deficiencies Medicines for severe nausea that cancer chemotherapy or pregnancy may cause Patient-controlled analgesia (PCA) for pain (this is IV ...

  8. Genève Reconnaissante

    CERN Multimedia

    2001-01-01

    Robert Cailliau (centre), with Geneva's Mayor Alain Vaissade (left) and Jean Erhardt, Secretary General of the Administrative Council of Geneva (right). Geneva recognised the contribution of two CERN people to the reputation of the city last Tuesday when Mayor Alain Vaissade presented the Genève Reconaissante Medal to Tim Berners-Lee and Robert Cailliau. Berners-Lee, who was not able to be present in person, invented the World Wide Web at CERN just over a decade ago, while Cailliau was his first collaborator. Quoting Cailliau, Vaissade said that whilst there is no doubt that something like the Web would have appeared sooner or later, the fact that it happened at CERN, in Geneva, was no accident. Both the Laboratory and the city are places where people from around the world meet and work in harmony.

  9. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  10. Options for helium circulation in a hydrogen production plant VHTR-Si: thermal-economic comparative

    International Nuclear Information System (INIS)

    Mendoza A, A.; Francois L, J. L.; Anaya D, A.

    2011-11-01

    The technologies that take advantage of the heat of nuclear reactors of IV generation are of great interest, due to their high energy efficiencies and to their strong economic potential. An example of these technologies is the sulfur-iodine process coupled to a nuclear reactor of high temperature cooled by helium. In this process heat is transferred from the nuclear reactor to the chemical plant by means of two cycles of helium interconnected by an intermediate heat exchanger. The first, denominated primary cycle of cooling, removes the heat of the nuclear reactor, transferring to the secondary cycle to be distributed to equipment s in the chemical plant. The pass of the helium gas through the equipment s that compose each one of the cycles, implies pressure losses that should be compensated necessarily by re-compression to maintain a stable state in the system, causing the energy consumption, usually rejected in the energy analyses. When to this energy is added the energy required in the hydrogen plant: energy required by the pumping systems, will decrease the efficiency of the nucleus-chemical complex, increasing the even cost of the hydrogen. In this work, three options to supply the compression energy and pumping (CEP) to the system are proposed, and these are analyzed thermodynamic and economically. The results indicate that to consider the CEP in the economic analysis increases between 1.5 and 3% the even cost of the hydrogen, and that the option with more energy efficiency is not necessarily the best for the nucleus-chemical complex. (Author)

  11. Report on generation IV technical working group 3 : liquid metal reactors

    International Nuclear Information System (INIS)

    Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

    2002-01-01

    This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

  12. Generation IV national program

    International Nuclear Information System (INIS)

    Preville, M.; Sadhankar, R.; Brady, D.

    2007-01-01

    This paper outlines the Generation IV National Program. This program involves evolutionary and innovative design with significantly higher efficiencies (∼50% compared to present ∼30%) - sustainable, economical, safe, reliable and proliferation resistant - for future energy security. The Generation IV Forum (GIF) effectively leverages the resources of the participants to meet these goals. Ten countries signed the GIF Charter in 2001

  13. Unleashing Gen Y: Marketing Mars to Millennials

    Science.gov (United States)

    Leahy, Bart D.; Hidalgo, Loretta; Kloberdanz, Cassie

    2007-01-01

    Space advocates need to engage Generation Y (born 1977-1999).This outreach is necessary to recruit the next generation of scientists and engineers to explore Mars. Space advocates in the non-profit, private, and government sectors need to use a combination of technical communication, marketing, and politics, to develop messages that resonate with Gen Y. Until now, space messages have been generated by and for college-educated white males; Gen Y is much more diverse, including as much as one third minorities. Young women, too, need to be reached. My research has shown that messages emphasizing technology, fun, humor, and opportunity are the best means of reaching the Gen Y audience of 60 million (US population is 300 million). The important things space advocates must avoid are talking down to this generation, making false promises, or expecting them to "wait their turn" before they can participate. This is the MTV generation! We need to find ways of engaging Gen Y now to build a future where human beings can live and work on the planet Mars. In addition to the messages themselves, advocates need to keep up with Gen Y' s social networking and use of iPods, cell phones, and the Internet. NASA and space advocacy groups can use these tools for "viral marketing," where young people share targeted space-related information via cell phones or the Internet because they like it. Overall, Gen Y is a socially dynamic and media-savvy group; advocates' space messages need to be sincere, creative, and placed in locations where Gen Y lives. Mars messages must be memorable!

  14. Neptunium (IV) oxalate solubility

    International Nuclear Information System (INIS)

    Luerkens, D.W.

    1983-07-01

    The equilibrium solubility of neptunium (IV) oxalate in nitric/oxalic acid solutions was determined at 22 0 C, 45 0 C, and 60 0 C. The concentrations of nitric/oxalic acid solutions represented a wide range of free oxalate ion concentration. A mathematical solubility model was developed which is based on the formation of the known complexes of neptunium (IV) oxalate. the solubility model uses a simplified concentration parameter which is proportional to the free oxalate ion concentration. The solubility model can be used to estimate the equilibrium solubility of neptunium (IV) oxalate over a wide range of oxalic and nitric acid concentrations at each temperature

  15. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2016.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  16. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2014.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  17. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2015.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  18. SAGE IV Pathfinder

    Data.gov (United States)

    National Aeronautics and Space Administration — Utilizing a unique, new occultation technique involving imaging, the SAGE IV concept will meet or exceed the quality of previous SAGE measurements at a small...

  19. Description of Guyruita gen. nov. and two new species (Ischnocolinae, Theraphosidae Descrição de Guyruita gen. nov. e duas novas espécies (Ischnocolinae, Theraphosidae

    Directory of Open Access Journals (Sweden)

    José P.L. Guadanucci

    2007-12-01

    Full Text Available The genus Guyruita gen. nov. and two new species from Brazil are described. Holothele waikoshiemi (Bertani & Araújo, 2005 from Venezuela is transferred here to the new genus. Guyruita gen. nov. differs from the remaining Ischnocolinae by the following features: labium densely occupied by a lot of cuspules (more than 100, intercheliceral intumescence absent, posterior sternal sigilla remote from margin, tarsal claws without teeth, tarsal scopula I-II undivided (tarsus II with a line of sparse setae, which does not divide the scopula, III-IV divided.É descrito o gênero Guyruita gen. nov. e duas espécies novas do Brasil. Holothele waikoshiemi (Bertani & Araújo, 2005 da Venezuela é transferido para o novo gênero. Guyruita gen. nov. difere dos outros Ischnocolinae pelas seguintes caracterísicas: lábio densamente ocupado por muitas cúspides (mais de 100, tumescência interqueliceral ausente, sigilla esternal posterior distante da margem, unhas tarsais sem dentes, escópula tarsal I e II inteiras (tarso II com uma fileira de cerdas esparsas, as quais não dividem a escópula, III e IV divididas.

  20. Algoritmos para genómica comparativa

    OpenAIRE

    Figueiras, Vasco da Rocha

    2010-01-01

    Com o surgimento da Genómica e da Proteómica, a Bioinformática conduziu a alguns dos avanços científicos mais relevantes do século XX. A Unidade de Investigação e Desenvolvimento do Biocant, parque biotecnológico de Cantanhede, assume actualmente o papel de motor no desenvolvimento da Genómica. O Biocant possui um importante sequenciador de larga escala que permite armazenar um elevado número de genomas, nomeadamente, genomas de bactérias. O estudo proposto reflecte a necessidade do Bio...

  1. Preserving Accuracy in GenBank

    DEFF Research Database (Denmark)

    Bidartondo, M.I.; Bruns, T. D.; Blackwell, M.

    2008-01-01

    GenBank, the public repository for nucleotide and protein sequences, is a critical resource for molecular biology, evolutionary biology, and ecology. While some attention has been drawn to sequence errors (1), common annotation errors also reduce the value of this database. In fact, for organisms...

  2. Teleport Generation 3 (Teleport Gen 3)

    Science.gov (United States)

    2016-03-01

    for high- throughput multi-band and multimedia connectivity from deployed locations to DISN and DoD Information Network (DoDIN) information sources and...2016 Major Automated Information System Annual Report Teleport Generation 3 (Teleport Gen 3) Defense Acquisition Management Information Retrieval...Program Information 4 Responsible Office 4 References 4 Program Description 5 Business Case 6 Program Status 8 Schedule 9

  3. Divergência genética entre genótipos de frangos tipo caipira

    Directory of Open Access Journals (Sweden)

    R. C. Veloso

    2015-10-01

    Full Text Available RESUMOObjetivou-se com este trabalho verificar a divergência genética entre sete genótipos de frangos tipo caipira da linhagem Redbro utilizando as características de desempenho por meio de técnicas de análise multivariada. Foram utilizados 840 pintos de um dia, machos, distribuídos em delineamento inteiramente ao acaso, dos seguintes genótipos: Caboclo, Carijó, Colorpak, Gigante Negro, Pesadão Vermelho, Pescoço Pelado e Tricolor. Após a consistência dos dados, foram avaliadas as seguintes variáveis: ganho em peso médio diário, consumo de ração médio diário e conversão alimentar, para os períodos: 1 a 28, 1 a 56, 1 a 70 e 1 a 84 dias de idade; peso corporal ao nascimento, aos 28, 56, 70 e aos 84 dias de idade. O desempenho dos genótipos foi avaliado por meio da análise de variância multivariada e da função discriminante linear de Fisher, usando os testes do maior autovalor de Roy e da união-interseção de Roy para as comparações múltiplas. O estudo da divergência genética foi feito por meio da análise por variáveis canônicas e pelo método de otimização de Tocher. Os genótipos Caboclo e Gigante Negro apresentaram médias canônicas diferentes dos demais genótipos. As duas primeiras variáveis canônicas explicaram 97,41% da variação entre os genótipos. A divergência genética entre os genótipos avaliados permitiu a formação de quatro grupos com os seguintes genótipos: grupo 1 - Colorpak; grupo 2 - Pesadão Vermelho e Pescoço Pelado; grupo 3 - Carijó e Tricolor; e grupo 4 - Caboclo e Gigante Negro.

  4. Development of multipurpose VHTR

    International Nuclear Information System (INIS)

    Ishikawa, Hiroshi; Aochi, Tetsuo; Hara, Masao

    1983-01-01

    In order to introduce atomic energy, which has been utilized mostly for electric power generation, into non-electric power field which amounts to 60 - 70% of energy demand in Japan, the development of a multi-purpose high temperature gas-cooled reactor has been advanced in the Japan Atomic Energy Research Institute. Including the progress and trend of the development of high temperature gas-cooled reactors in foreign countries, the features, necessity, the state of research and development and the way of thinking about heat utilization system regarding the reactors of this type are summarized. Since the Dragon Project of OECD in 1959, the course of the development of high temperature gas-cooled reactors is described. In Japan, the utilization of nuclear thermal energy for iron-making process was investigated to resolve environmental problems and to get rid of coal. It was decided to construct an experimental reactor, aiming at the start of operation around 1990. The features of high temperature gas-cooled reactors, the utilization mode of nuclear thermal energy, the design of an experimental reactor, the research and development related to the experimental reactor and the heat utilization system for the experimental reactor, the trend of development in FRG, USA and USSR are described. (Kako, I.)

  5. Next Gen One Portal Usability Evaluation

    Science.gov (United States)

    Cross, E. V., III; Perera, J. S.; Hanson, A. M.; English, K.; Vu, L.; Amonette, W.

    2018-01-01

    Each exercise device on the International Space Station (ISS) has a unique, customized software system interface with unique layouts / hierarchy, and operational principles that require significant crew training. Furthermore, the software programs are not adaptable and provide no real-time feedback or motivation to enhance the exercise experience and/or prevent injuries. Additionally, the graphical user interfaces (GUI) of these systems present information through multiple layers resulting in difficulty navigating to the desired screens and functions. These limitations of current exercise device GUI's lead to increased crew time spent on initiating, loading, performing exercises, logging data and exiting the system. To address these limitations a Next Generation One Portal (NextGen One Portal) Crew Countermeasure System (CMS) was developed, which utilizes the latest industry guidelines in GUI designs to provide an intuitive ease of use approach (i.e., 80% of the functionality gained within 5-10 minutes of initial use without/limited formal training required). This is accomplished by providing a consistent interface using common software to reduce crew training, increase efficiency & user satisfaction while also reducing development & maintenance costs. Results from the usability evaluations showed the NextGen One Portal UI having greater efficiency, learnability, memorability, usability and overall user experience than the current Advanced Resistive Exercise Device (ARED) UI used by astronauts on ISS. Specifically, the design of the One-Portal UI as an app interface similar to those found on the Apple and Google's App Store, assisted many of the participants in grasping the concepts of the interface with minimum training. Although the NextGen One-Portal UI was shown to be an overall better interface, observations by the test facilitators noted specific exercise tasks appeared to have a significant impact on the NextGen One-Portal UI efficiency. Future updates to

  6. The European gen-set market: growth and consolidation mean joy and pain

    International Nuclear Information System (INIS)

    French, Ian

    2000-01-01

    The changes in the European gen-set market are discussed. In recent years the market has undergone a period of increasing consolidation: prices fell and some companies folded. However, the market is not dead and continued growth is expected over the next five years although the compound rate is forecast to be only 1.5%. The article is presented under the sub-headings of (i) current market situation; (ii) product lifecycle; (iii) shipments by technology; (iv) market deregulation; (v) technology overview (spark ignition, compression ignition and gas turbines) (vi) European market: national overview and (vii) key market challenges (competition, emissions and over capacity)

  7. IV access in dental practice.

    LENUS (Irish Health Repository)

    Fitzpatrick, J J

    2009-04-01

    Intravenous (IV) access is a valuable skill for dental practitioners in emergency situations and in IV sedation. However, many people feel some apprehension about performing this procedure. This article explains the basic principles behind IV access, and the relevant anatomy and physiology, as well as giving a step-by-step guide to placing an IV cannula.

  8. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning

  9. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

  10. Internet Economics IV

    Science.gov (United States)

    2004-08-01

    edts.): Internet Economics IV Technical Report No. 2004-04, August 2004 Information Systems Laboratory IIS, Departement of Computer Science University of...level agreements (SLA), Information technology (IT), Internet address, Internet service provider 16. PRICE CODE 17. SECURITY CLASSIFICATION 18... technology and its economic impacts in the Internet world today. The second talk addresses the area of AAA protocol, summarizing authentication

  11. Uranium (IV) carboxylates - I

    Energy Technology Data Exchange (ETDEWEB)

    Satpathy, K C; Patnaik, A K [Sambalpur Univ. (India). Dept. of Chemistry

    1975-11-01

    A few uranium(IV) carboxylates with monochloro and trichloro acetic acid, glycine, malic, citric, adipic, o-toluic, anthranilic and salicylic acids have been prepared by photolytic methods. The I.R. spectra of these compounds are recorded and basing on the spectral data, structure of the compounds have been suggested.

  12. PLATO IV Accountancy Index.

    Science.gov (United States)

    Pondy, Dorothy, Comp.

    The catalog was compiled to assist instructors in planning community college and university curricula using the 48 computer-assisted accountancy lessons available on PLATO IV (Programmed Logic for Automatic Teaching Operation) for first semester accounting courses. It contains information on lesson access, lists of acceptable abbreviations for…

  13. A Study on the Planning of Technology Development and Research for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Moon Hee; Kim, H. R.; Kim, H. J. and others

    2005-08-15

    This study aimed at the planning the domestic technology development of the Gen IV and the formulating the international collaborative project contents and executive plan for 'A Validity Assessment and Policies of the R and D of Generation IV Nuclear Energy Systems'. The results of the study include follows; - Survey of the technology state in the fields of the Gen IV system specific technologies and the common technologies, and the plans of the international collaborative research - Drawing up the executive research and development plan by the experts of the relevant technology field for the systems which Korean will participate in. - Formulating the effective conduction plan of the program reflecting the view of the experts from the industry, the university and the research institute. - Establishing the plan for estimation of the research fund and the manpower for the efficient utilization of the domestic available resources. This study can be useful material for evaluating the appropriateness of the Korea's participation in the international collaborative development of the Gen IV, and can be valuably utilized to establish the strategy for the effective conduction of the program. The executive plan of the research and development which was produced in this study will be used to the basic materials for the establishing the guiding direction and the strategic conduction of the program when the research and development is launched in the future.

  14. Genética humana e sociedade

    OpenAIRE

    Rosa, Vivian Leyser da

    2000-01-01

    Tese (doutorado) - Universidade Federal de Santa Catarina, Centro de Ciências da Educação. Análise do campo de estudos sobre o entendimento público da ciência, distinguindo os modelos de deficit cognitivo e interativo, bem como suas implicações na esfera educacional. Estudo do panorama dos avanços atuais da genética humana, do ponto de vista científico, ético e social. Análise de aspectos relativos ao ensino de genética humana nos cursos de graduação da área da saúde, em nove Universidades...

  15. A Review of Alloy 800H for Applications in the Gen IV Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Ren, Weiju; Swindeman, Robert W.

    2010-01-01

    Alloy 800H is currently under consideration for applications in the Next Generation Nuclear Plant at operational temperatures above 750 C. To provide supporting information in this paper at the attempt to facilitate the consideration, service requirements of the nuclear system for structural materials is first described; and then an extensive review of Alloy 800H is given on its codification with respect to development and research history, mechanical behavior and design allowables, metallurgical aging resistance, environmental effect considerations, data requirements and availability, weldments, as well as many other aspects relevant to the intended nuclear application; an finally further research and development activities to support the materials qualification are suggested.

  16. Project planning of Gen-IV sodium cooled fast reactor technology

    International Nuclear Information System (INIS)

    Yoo, Jaewoon; Joo, H. K.; Cho, C. H.; Kim, Y. G.; Lee, D. U.; Jin, M. W.

    2013-05-01

    The project program will be established to shorten the design schedule by sharing the design man power and experimental facility, and by introducing the proven technology through international collaboration and the project plan including preliminary specific design, technology validation and fuel design validation plan will be more detail by reviewing the plan at the International Technical Review Meeting (ITRM). Periodic project progress review meeting will be held to find the technical issues and to resolve them. The results of the progress review meeting will be reflected into the final assessment of research project. The project progress review meeting will be held every quarter and external expert will also participate in the meeting. In parallel with the PGSFR development, innovative small modular SFR will be developed aiming to the international nuclear market. The system and component technologies of both system can be shared but innovative concept will be implemented into the design. Ultra long life core design concept and supercritical CO 2 Brayton cycle will be considered as the innovative concept for enhancing the plant economy and safety

  17. Database of thermophysical properties of liquid metal coolants for GEN-IV

    International Nuclear Information System (INIS)

    Sobolev, V.

    2011-01-01

    The report presents an Fe-Cr interatomic potential to model high-Cr ferritic alloys. The potential is fitted to thermodynamic and point-defect properties obtained from density functional theory (DFT) calculations and experiments. The developed potential is also benchmarked against other potentials available in literature. It shows particularly good agreement with the DFT obtained mixing enthalpy of the random alloy, the formation energy of intermetallics and experimental excess vibrational entropy and phase diagram. In addition, DFT calculated point-defect properties, both interstitial and substitutional, are well reproduced, as is the screw dislocation core structure. As a first validation of the potential, we study the precipitation hardening of Fe-Cr alloys via static simulations of the interaction between Cr precipitates and screw dislocations. It is concluded that the description of the dislocation core modification near a precipitate might have a significant influence on the interaction mechanisms observed in dynamic simulations.

  18. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    Energy Technology Data Exchange (ETDEWEB)

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric [Commissariat a l' Energie Atomique (C.E.A.), Direction de l' Energie Nucleaire, Centre d' Etudes de Cadarache, 13108 Saint Paul lez Durance Cedex (France)

    2008-07-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC{sup R} process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  19. Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Wosko, Paul; Sundram, S. K.

    2012-10-16

    New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.

  20. Project planning of Gen-IV sodium cooled fast reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jaewoon; Joo, H. K.; Cho, C. H.; Kim, Y. G.; Lee, D. U.; Jin, M. W.

    2013-05-15

    The project program will be established to shorten the design schedule by sharing the design man power and experimental facility, and by introducing the proven technology through international collaboration and the project plan including preliminary specific design, technology validation and fuel design validation plan will be more detail by reviewing the plan at the International Technical Review Meeting (ITRM). Periodic project progress review meeting will be held to find the technical issues and to resolve them. The results of the progress review meeting will be reflected into the final assessment of research project. The project progress review meeting will be held every quarter and external expert will also participate in the meeting. In parallel with the PGSFR development, innovative small modular SFR will be developed aiming to the international nuclear market. The system and component technologies of both system can be shared but innovative concept will be implemented into the design. Ultra long life core design concept and supercritical CO{sub 2} Brayton cycle will be considered as the innovative concept for enhancing the plant economy and safety.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  2. Electromagnet Response Time Tests on Primary CRDM of a Prototype Gen-IV SFR

    International Nuclear Information System (INIS)

    Lee, Jae-Han; Koo, Gyeong-Hoi

    2015-01-01

    This paper identifies the electromagnetic response characteristics of the electromagnet of a primary control rod drive mechanism (CRDM) used for the reactor scram function. The test measures the electromagnet response time required to release an armature from a stator controlled by a loss of an electromagnetic force on an armature after shorting a power supply to an electromagnet coil. These tests are carried out while changing the electromagnet core material, an assist spring, and an armature holding current. The main factors influencing the test parameters on the response are found to be the armature holding current for holding the armature loads, and the material type of the electromagnet cores. The minimum response time is 0.13 seconds in the case of using SS410 material as an armature, while the S10C material as an armature has a response time of 0.21 seconds. Electromagnet response time characteristics from the test results will be evaluated by comparing the precise moving data of an electromagnet armature through the use of a high-speed camera and a potentiometer in the future

  3. Tritium Sequestration in Gen IV NGNP Gas Stream via Proton Conducting Ceramic Pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Fanglin Frank [Univ. of South Carolina, Columbia, SC (United States); Adams, Thad M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brinkman, Kyle [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reifsnider, Kenneth [Univ. of South Carolina, Columbia, SC (United States)

    2011-09-30

    Several types of high-temperature proton conductors based on SrCeO3 and BaCeO3 have been systematically investigated in this project for tritium separation in NGNP applications. One obstacle for the field application is the chemical stability issues in the presence of steam and CO2 for these proton conductors. Several strategies to overcome such issues have been evaluated, including A site doping and B site co-doping method for perovskite-structured proton conductors. Novel zirconium-free proton conductors have also been developed with improved electrical conductivity and enhanced chemical stability. Novel catalytic materials for the proton-conducting separation membranes have been investigated. A tubular geometry proton-conducting membrane has been developed for the proton separation membranes. Total dose rate estimated from tritium decay (beta emission) under realistic membrane operating conditions, combined with electron irradiation experiments, indicates that proton ceramic materials possess the appropriate radiation stability for this application.

  4. Studies on woloszynskioid dinoflagellates IV: the genus Biecheleria gen. nov

    DEFF Research Database (Denmark)

    Moestrup, Øjvind; Lindberg, Karin; Daugbjerg, Niels

    2009-01-01

    in the dinoflagellates. Biecheleria also comprises the brackish water species Biecheleria baltica sp. nov. (presently identified as Woloszynskia halophila) and the marine species Biecheleria natalensis (syn. Gymnodinium natalense). Gymnodinium halophilum described in 1952 by B. Biecheler but apparently not subsequently...... refound, is transferred to Biecheleria. The Suessiaceae further includes the marine species Protodinium simplex, described by Lohmann in 1908 but shortly afterwards (1921) transferred to Gymnodinium by Kofoid and Swezy and subsequently known as Gymnodinium simplex. It only distantly related to Gymnodinium...

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

    International Nuclear Information System (INIS)

    Choi, Chiwoong; Jeong, Taekyung; Lee, Kwilim; Jeong, Jaeho; Ha, Kwiseok

    2015-01-01

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

  6. Training Courses in Support of GEN-IV Development – The Case of SVBR Technology

    International Nuclear Information System (INIS)

    Kondaurov, A.; Zaitseva, N.; Yunikova, A.; Artisiuk, V.

    2014-01-01

    Conclusions: For prototype nuclear power reactor the development of training materials requires high level expertise from the R&D side. The First International Course focusing the SVBR technology was developed and piloted in ROSATOM Central Institute for Continuing Education&Training to support HRD for Open Joint-Stock Company «AKME-engineering» - owner and operator of SVBR-100. The Course is available for international participants

  7. Minor actinides transmutation potential: state of art for GEN IV sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Buiron, Laurent

    2015-01-01

    In the frame of the R and D program relative to the 1991 French act on nuclear waste management, fast neutron systems have shown relevant characteristics that meet both requirements on sustainable resources management and waste minimization. They also offer flexibility by mean of burner or breeder configurations allowing mastering plutonium inventory without significant impact on core safety. From the technological point of view, sodium cooled fast reactor are considered in order to achieve mean term industrial deployment. The present document summaries the main results of R and D program on minor actinides transmutation in sodium fast reactor since 2006 following recommendation of the first part of the 1991 French act. Both homogeneous and heterogeneous management achievable performances are presented for 'evolutionary' SFR V2B core as well as low void worth CFV core for industrial scale configurations (1500 MWe). Minor actinides transmutation could be demonstrated in the ASTRID reactor with the following configurations: - a 2%vol Americium content for the homogeneous mode, - a 10%vol Americium content for the heterogeneous mode, without any substantial modification of the main core safety parameters and only limited impacts on the associated fuel cycle (manufacturing issues are not considered here). In order to achieve such goal, a wide range of experimental irradiations driven by transmutation scenarios have to be performed for both homogeneous and heterogeneous minor actinides management. (author) [fr

  8. Optimized, Competitive Supercritical-CO2 Cycle GFR for Gen IV Service

    International Nuclear Information System (INIS)

    M.J. Driscoll; P. Hejzlar; G. Apostolakis

    2008-01-01

    An overall plant design was developed for a gas-cooled fast reactor employing a direct supercritical Brayton power conversion system. The most important findings were that (1) the concept could be capital-cost competitive, but startup fuel cycle costs are penalized by the low core power density, specified in large part to satisfy the goal of significant post-accident passive natural convection cooling; (2) active decay heat removal is preferable as the first line of defense, with passive performance in a backup role; (3) an innovative tube-in-duct fuel assembly, vented to the primary coolant, appears to be practicable; and (4) use of the S-Co2 GFR to support hydrogen production is a synergistic application, since sufficient energy can be recuperated from the product H2 and 02 to allow the electrolysis cell to run 250 C hotter than the reactor coolant, and the water boilers can be used for reactor decay heat removal. Increasing core power density is identified as the top priority for future work on GFRs of this type

  9. Tritium Sequestration in Gen IV NGNP Gas Stream via Proton Conducting Ceramic Pumps

    International Nuclear Information System (INIS)

    Chen, Franglin Frank; Adams, Thad M.; Brinkman, Kyle; Reifsnider, Kenneth

    2011-01-01

    Several perovskite structured proton conductors based on SrCeO 3 and BaCeO 3 have been investigated in the project. The solid solutions for SrCeO 3 and BaCeO 3 were first investigated. The morphological and electrical properties of Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ with x varying from 0 to 1 prepared by a modified Pechini method were investigated as potential high temperature proton conductors. Dense microstructures were achieved for all the samples upon sintering at 1500ees)C for 5 h. The phase structure analysis indicated that perovskite phase was formed for 0≤x≤0.2, while for x larger than 0.5, impurity phases of Sr 2 CeO 4 and Y 2 O 3 appeared. The stability tests indicated that the resistance to boiling water for Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ was between that of BaCe 0.8 Y 0.2 O 3-δ and SrCe 0.8 Y 0.2 O 3-δ Due to the tendency of the reaction with CO 2 for both BaCe 0.8 Y 0.2 O 3-δ and SrCe 0.8 Y 0.2 O 3-δ , it was not surprising that Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ was also not stable in CO 2 containing atmospheres. The conductivity tests indicated that Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ possessed the electrical conductivity between BaCe 0.8 Y 0.2 O 3-δ and SrCe 0.8 Y 0.2 O 3-δ . The conductivity decreased and the activation energy increased with the increase in Sr content in Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ .

  10. Preliminary Comparative Evaluation Study on Reference Design of GEN-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Yoon Sub; Kim, Yeong Il; Hong, Ser Gi (and others)

    2005-11-15

    A fast reactor has a good transmutation capability and it enables breeding of fuel and use of a closed fuel cycle. By these characteristics of a fast reactor, the limited uranium resources of the world can be much more effectively utilized and the nuclear wastes of a high level of radioactivity and toxicity from the current nuclear power reactors of LWRs and HWRs can be drastically reduced in its volume and the management of the wastes can be easily treated. Also electricity can be generated more effectively since a fast reactor has the feature of high operation temperature. These features of a fast reactor makes it inevitable on a long term basis to construct fast reactors in Korea. The domestic fast reactor technology level, however, is at the level of coming out of a beginning stage and needs utilization of international expertise. Recently an international cooperation program called GIF has been formulated and our KALIMER was selected as one of the two reference designs for the international joint R and D works with JSFR of Japan. In the current frame of the GIF program, the two selected reference designs are supposed to be evaluated against each other in future and one design is to be finally selected. To make the international cooperation program directed more useful to our fast reactor technology development, it is required to strengthen the competitiveness of KALIMER so that it can be selected. To meet the necessity, a study was made in this research for pre-evaluation of the GIF reference designs and setting up plans for development of designs and technology that will enhance the competitiveness of KALIMER.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  12. Enhanced Design Alternative IV

    International Nuclear Information System (INIS)

    Kramer, N.E.

    1999-01-01

    This report evaluates Enhanced Design Alternative (EDA) IV as part of the second phase of the License Application Design Selection (LADS) effort. The EDA IV concept was compared to the VA reference design using criteria from the Design Input Request for LADS Phase II EDA Evaluations (CRWMS M and O 1999b) and (CRWMS M and O 1999f). Briefly, the EDA IV concept arranges the waste packages close together in an emplacement configuration known as line load. Continuous pre-closure ventilation keeps the waste packages from exceeding their 350 C cladding and 200 C (4.3.6) drift wall temperature limits. This EDA concept keeps relatively high, uniform emplacement drift temperatures (post-closure) to drive water away from the repository and thus dry out the pillars between emplacement drifts. The waste package is shielded to permit human access to emplacement drifts and includes an integral filler inside the package to reduce the amount of water that can contact the waste form. Closure of the repository is desired 50 years after first waste is emplaced. Both backfill and drip shields will be emplaced at closure to improve post-closure performance. The EDA IV concept includes more defense-in-depth layers than the VA reference design because of its backfill, drip shield, waste package shielding, and integral filler features. These features contribute to the low dose-rate to the public achieved during the first 10,000 years of repository life as shown in Figure 3. Investigation of the EDA IV concept has led to the following general conclusions: (1) The total life cycle cost for EDA IV is about $21.7 billion which equates to a $11.3 billion net present value (both figures rounded up). (2) The incidence of design basis events for EDA IV is similar to the VA reference design. (3) The emplacement of the waste packages in drifts will be similar to the VA reference design. However, heavier equipment may be required because the shielded waste package will be heavier. (4) The heavier

  13. Review on Korea Participation of Generation IV International Forum (GIF)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jewhan; Jeong, Ji-Young; Hahn, Dohee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Generation IV International Forum (GIF) originates from US proposal of an initiative in 2000. The vision was to leapfrog LWR technology and collaborate with international partners to share R and D on advanced nuclear systems. Nine countries and EU joined the initiative and Gen IV concept was defined via technology goals and legal framework. Two years study with more than 100 experts worldwide has evaluated nearly 100 reactor designs and down selected six most promising concepts. In 2005, the first signatures on Framework Agreement were collected and the first research projects were defined in 2006. Korea is one of the founding members of GIF and actively participating in various areas. In 2013, TD was assigned to Korean expert and Korea is endeavoring to enhance the benefit of participation since this turning point. In this paper, pros and cons of engaging with GIF were briefly introduced and items to maximize the benefit were suggested.

  14. Review on Korea Participation of Generation IV International Forum (GIF)

    International Nuclear Information System (INIS)

    Lee, Jewhan; Jeong, Ji-Young; Hahn, Dohee

    2015-01-01

    Generation IV International Forum (GIF) originates from US proposal of an initiative in 2000. The vision was to leapfrog LWR technology and collaborate with international partners to share R and D on advanced nuclear systems. Nine countries and EU joined the initiative and Gen IV concept was defined via technology goals and legal framework. Two years study with more than 100 experts worldwide has evaluated nearly 100 reactor designs and down selected six most promising concepts. In 2005, the first signatures on Framework Agreement were collected and the first research projects were defined in 2006. Korea is one of the founding members of GIF and actively participating in various areas. In 2013, TD was assigned to Korean expert and Korea is endeavoring to enhance the benefit of participation since this turning point. In this paper, pros and cons of engaging with GIF were briefly introduced and items to maximize the benefit were suggested

  15. Targeted NextGen Capabilities for 2025

    Science.gov (United States)

    2011-11-01

    increased arrival capacity to single runways by reducing longitudinal wake separation standards for Instrument Flight Rules ( IFR ) operations under certain...b. ABSTRACT unclassified c. THIS PAGE unclassified Standard Form 298 (Rev. 8-98) Prescribed by ANSI Std Z39-18 Targeted NextGen Capabilities...The examples cited are not intended to cover every aircraft and every flight. In some instances, the available capabilities for 2025 will not be

  16. Development of C/C composite for the core component of the high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H

    2005-01-15

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite.

  17. Development of C/C composite for the core component of the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H.

    2005-01-01

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite

  18. A sputnik IV saga

    Science.gov (United States)

    Lundquist, Charles A.

    2009-12-01

    The Sputnik IV launch occurred on May 15, 1960. On May 19, an attempt to deorbit a 'space cabin' failed and the cabin went into a higher orbit. The orbit of the cabin was monitored and Moonwatch volunteer satellite tracking teams were alerted to watch for the vehicle demise. On September 5, 1962, several team members from Milwaukee, Wisconsin made observations starting at 4:49 a.m. of a fireball following the predicted orbit of Sputnik IV. Requests went out to report any objects found under the fireball path. An early morning police patrol in Manitowoc had noticed a metal object on a street and had moved it to the curb. Later the officers recovered the object and had it dropped off at the Milwaukee Journal. The Moonwarch team got the object and reported the situation to Moonwatch Headquarters at the Smithsonian Astrophysical Observatory. A team member flew to Cambridge with the object. It was a solid, 9.49 kg piece of steel with a slag-like layer attached to it. Subsequent analyses showed that it contained radioactive nuclei produced by cosmic ray exposure in space. The scientists at the Observatory quickly recognized that measurements of its induced radioactivity could serve as a calibration for similar measurements of recently fallen nickel-iron meteorites. Concurrently, the Observatory directorate informed government agencies that a fragment from Sputnik IV had been recovered. Coincidently, a debate in the UN Committee on Peaceful Uses of Outer Space involved the issue of liability for damage caused by falling satellite fragments. On September 12, the Observatory delivered the bulk of the fragment to the US Delegation to the UN. Two days later, the fragment was used by US Ambassador Francis Plimpton as an exhibit that the time had come to agree on liability for damage from satellite debris. He offered the Sputnik IV fragment to USSR Ambassador P.D. Morozov, who refused the offer. On October 23, Drs. Alla Massevitch and E.K. Federov of the USSR visited the

  19. Recommendations and Requirements for GenCade Simluations

    Science.gov (United States)

    2014-08-01

    will report whether or not GenCade is enabled. If GenCade is disabled , the user will need a new license that includes GenCade...any depth but usually are not deeper than the seaward edge of the surf - zone. In the same way that some shorelines are less desirable for use in...Conference, 1919–1937. ASCE. Wang, P., N. C. Kraus, and R. A. Davis. 1998. Total rate of longshore sediment transport in the surf zone: Field

  20. O impacto da genética na asma infantil

    OpenAIRE

    Pinto,Leonardo A.; Stein,Renato T.; Kabesch,Michael

    2008-01-01

    OBJETIVO: Apresentar os resultados dos estudos mais importantes e recentes sobre a genética da asma. Estes dados devem auxiliar os clínicos gerais a compreender o impacto da genética sobre este distúrbio complexo e como os genes e polimorfismos influenciam a asma e a atopia. FONTES DOS DADOS: Os dados foram coletados do banco de dados MEDLINE. Os estudos de associação genética foram selecionados do Genetic Association Database, um repositório de estudos de associação genética de doenças e dis...

  1. Aconselhamento genético Genetic counseling

    Directory of Open Access Journals (Sweden)

    João Monteiro de Pina-Neto

    2008-08-01

    Full Text Available OBJETIVO: Esta revisão sobre aconselhamento genético (AG teve o objetivo de mostrar os conceitos atuais e os princípios filosóficos e éticos aceitos na grande maioria dos países e recomendados pela Organização Mundial da Saúde, as fases do processo, seus resultados e o impacto psicológico de uma doença genética em uma família. FONTES DOS DADOS: Os conceitos apresentados são baseados em uma síntese histórica da literatura sobre AG desde a década de 1930 até o momento atual, sendo que os artigos citados representam os principais trabalhos publicados e que hoje fundamentam a teoria e a prática do AG. SÍNTESE DOS DADOS: O AG modernamente é definido como um processo de comunicação que trata dos problemas humanos relacionados à ocorrência de uma doença genética em uma família. É fundamental que os profissionais da saúde conheçam os aspectos psicológicos desencadeados pela doença genética e como estes aspectos podem ser manejados. Vivemos ainda na genética humana e médica uma fase de predomínio dos aspectos técnicos e científicos e de pouca ênfase no estudo das reações emocionais e dos processos de adaptação das pessoas a estas doenças, o que leva ao baixo entendimento dos clientes sobre os fatos ocorridos, com conseqüências negativas sobre a vida familiar e para a sociedade. CONCLUSÕES: Conclui-se pela necessidade de que as famílias com doenças genéticas sejam encaminhadas para AG e que os profissionais desta área invistam mais na humanização do atendimento, desenvolvendo mais as técnicas do AG psicológico não-diretivo.OBJECTIVE: The objective of this review of genetic counseling (GC is to describe the current concepts and philosophical and ethical principles accepted by the great majority of countries and recommended by the World Health Organization, the stages of the process, its results and the psychological impact that a genetic disease has on a family. SOURCES: The concepts presented are

  2. TidGen Power System Commercialization Project

    Energy Technology Data Exchange (ETDEWEB)

    Sauer, Christopher R. [President & CEO; McEntee, Jarlath [VP Engineering & CTO

    2013-12-30

    ORPC Maine, LLC, a wholly-owned subsidiary of Ocean Renewable Power Company, LLC (collectively ORPC), submits this Final Technical Report for the TidGen® Power System Commercialization Project (Project), partially funded by the U.S. Department of Energy (DE-EE0003647). The Project was built and operated in compliance with the Federal Energy Regulatory Commission (FERC) pilot project license (P-12711) and other permits and approvals needed for the Project. This report documents the methodologies, activities and results of the various phases of the Project, including design, engineering, procurement, assembly, installation, operation, licensing, environmental monitoring, retrieval, maintenance and repair. The Project represents a significant achievement for the renewable energy portfolio of the U.S. in general, and for the U.S. marine hydrokinetic (MHK) industry in particular. The stated Project goal was to advance, demonstrate and accelerate deployment and commercialization of ORPC’s tidal-current based hydrokinetic power generation system, including the energy extraction and conversion technology, associated power electronics, and interconnection equipment capable of reliably delivering electricity to the domestic power grid. ORPC achieved this goal by designing, building and operating the TidGen® Power System in 2012 and becoming the first federally licensed hydrokinetic tidal energy project to deliver electricity to a power grid under a power purchase agreement in North America. Located in Cobscook Bay between Eastport and Lubec, Maine, the TidGen® Power System was connected to the Bangor Hydro Electric utility grid at an on-shore station in North Lubec on September 13, 2012. ORPC obtained a FERC pilot project license for the Project on February 12, 2012 and the first Maine Department of Environmental Protection General Permit issued for a tidal energy project on January 31, 2012. In addition, ORPC entered into a 20-year agreement with Bangor Hydro Electric

  3. A Study on planning of promotion for international collaborative development of Generation IV Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Hee, Chang Moon; Yang, M. S.; Ha, J. J.

    2006-06-01

    Korea has participated in the international collaboration programs for the development of future nuclear energy systems driven by the countries holding advanced nuclear technology and Korea and U. S. have cooperated in the INERI. This study is mainly at developing the plan for participation in the collaborative development of the Gen IV, searching the participation strategy for INERI and the INPRO, and the international cooperation in these programs. Contents and scope of the study for successful achievement are as follows; - Investigation and analysis of international and domestic trends related to advanced nuclear technologies - Development of the plan for collaborative development of the Gen IV and conducting the international cooperation activities - Support for the activities related to I-NERI between Korea and U. S. and conducting the international cooperation - International cooperation activities for the INPRO This study can be useful for planning the research plan and setting up of the strategy of integrating the results of the international collaboration and the domestic R and D results by combining the Gen IV and the domestic R and D in the field of future nuclear technology. Furthermore, this study can contribute to establishing the effective foundation and broadening the cooperation activities not only with the advanced countries for acquisition of the advanced technologies but also with the developing countries for the export of the domestic nuclear energy systems

  4. Analysis of a Spanish energy scenario with Generation IV nuclear reactors

    International Nuclear Information System (INIS)

    Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

    2013-01-01

    Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

  5. Overview of the CEA R and D support to generation IV nuclear energy systems

    International Nuclear Information System (INIS)

    Carre, Frank; Anzieu, Pascal; Billot, Philippe; Brossard, Philippe; Fiorini, Gian-Luigi

    2004-01-01

    As a result of an early technology road-map performed at the end of 2000, the CEA selected a sequenced development of advanced gas cooled high temperature nuclear systems as main focus for its R and D programme on future nuclear energy systems. The selection of this research objectives originates both from the significance of fast neutrons and high temperature for nuclear energy to meet the needs anticipated beyond 2020/2030, and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR likely to meet international market needs around 2020. The second step is a Very High Temperature Reactor (> 950 deg. C) to efficiently produce, among others, hydrogen though thermo-chemical water splitting or to generate electricity with an efficiency above 50%. The third step of the Path is a Gas Fast Reactor that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct-cycle helium turbine for electricity production and full recycle of actinides. The paper succinctly presents the R and D program launched in 2001 by the CEA with industrial partners on the 'Gas Technology Path', which is destined to become the contribution of France to the development of the VHTR and the GFR within the next phase of the Generation IV Forum

  6. Overview of the CEA R and D support to generation IV nuclear energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Carre, Frank; Anzieu, Pascal; Billot, Philippe; Brossard, Philippe; Fiorini, Gian-Luigi

    2004-07-01

    As a result of an early technology road-map performed at the end of 2000, the CEA selected a sequenced development of advanced gas cooled high temperature nuclear systems as main focus for its R and D programme on future nuclear energy systems. The selection of this research objectives originates both from the significance of fast neutrons and high temperature for nuclear energy to meet the needs anticipated beyond 2020/2030, and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR likely to meet international market needs around 2020. The second step is a Very High Temperature Reactor (> 950 deg. C) to efficiently produce, among others, hydrogen though thermo-chemical water splitting or to generate electricity with an efficiency above 50%. The third step of the Path is a Gas Fast Reactor that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct-cycle helium turbine for electricity production and full recycle of actinides. The paper succinctly presents the R and D program launched in 2001 by the CEA with industrial partners on the 'Gas Technology Path', which is destined to become the contribution of France to the development of the VHTR and the GFR within the next phase of the Generation IV Forum.

  7. Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

  8. Hepatic imaging in stage IV-S neuroblastoma

    International Nuclear Information System (INIS)

    Franken, E.A. Jr.; Smith, W.L.; Iowa Univ., Iowa City; Cohen, M.D.; Kisker, C.T.; Platz, C.E.

    1986-01-01

    Stage IV-S neuroblastoma describes a group of infants with tumor spread limited to liver, skin, or bone marrow. Such patients, who constitute about 25% of affected infants with neuroblastoma, may expect spontaneous tumor remission. We report 18 infants with Stage IV-S neuroblastoma, 83% of whom had liver involvement. Imaging investigations included Technetium 99m sulfur colloid scan, ultrasound, and CT. Two patterns of liver metastasis were noted: ill-defined nodules or diffuse tumor throughout the liver. Distinction of normal and abnormal liver with diffuse type metastasis could be quite difficult, particularly with liver scans. We conclude that patients with Stage IV-S neuroblastoma have ultrasound or CT examination as an initial workup, with nuclear medicine scans reserved for followup studies. (orig.)

  9. Diaquatetrabromidotin(IV trihydrate

    Directory of Open Access Journals (Sweden)

    Fei Ye

    2012-09-01

    Full Text Available The title compound, [SnBr4(H2O2]·3H2O, forms large colourless crystals in originally sealed samples of tin tetrabromide. It constitutes the first structurally characterized hydrate of SnBr4 and is isostructural with the corresponding hydrate of SnCl4. It is composed of SnIV atoms octahedrally coordinated by four Br atoms and two cis-related water molecules. The octahedra exhibit site symmetry 2. They are arranged into columns along [001] via medium–strong O—H...O hydrogen bonds involving the two lattice water molecules (one situated on a twofold rotation axis while the chains are interconnected via longer O—H...Br hydrogen bonds, forming a three-dimensional network.

  10. Cyclopentadienyluranium(IV) acetylacetonates

    International Nuclear Information System (INIS)

    Bagnall, K.W.; Edwards, J.; Rickard, C.E.F.; Tempest, A.C.

    1979-01-01

    Cyclopentadienyluranium(IV) acetylacetonate complexes, (eta 5 C 5 H 5 )UClsub(3-x)(acac)sub(x), where x = 1 or 2, and the corresponding bis triphenylphosphine oxide (tppo) complexes have been prepared. The bis cyclopentadienyl complexes, (eta 5 C 5 H 5 ) 2 U(acac) 2 and (eta 5 C 5 H 5 ) 2 UCl(acac)(tppo) 2 have also been prepared and are stable with respect to disproportionation, whereas (eta 5 C 5 H 5 ) 2 UCl(acac) is not. The IR and UV/visible spectra of the complexes are reported, together with some additional information on the UCl 2 (acac) 2 thf and -tppo systems. (author)

  11. Towards an International Culture: Gen Y Students and SNS?

    Science.gov (United States)

    Lichy, Jessica

    2012-01-01

    This article reports the findings of a small-scale investigation into the Internet user behaviour of generation Y (Gen Y) students, with particular reference to social networking sites. The study adds to the literature on cross-cultural Internet user behaviour with specific reference to Gen Y and social networking. It compares how a cohort of…

  12. R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

    2005-01-01

    The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

  13. Caracterização genético-clínica de pacientes com fenilcetonúria no Estado de Alagoas = Genetic and clinical characterization of patients with phenylketonuria in Alagoas state, Brazil

    Directory of Open Access Journals (Sweden)

    Santos, Emerson Santana

    2012-01-01

    Conclusões: O genótipo V388M/IVS10nt11G>A foi o mais prevalente. Trinta por cento dos pacientes foram sintomáticos, provavelmente pela natureza das mutações, não adesão ao tratamento, tratamento inadequado e/ou diagnóstico tardio

  14. Variabilidad genética de Plasmodium falciparum en pacientes con malaria grave y malaria no complicada en Iquitos - Perú

    Directory of Open Access Journals (Sweden)

    Gisely Hijar G

    2002-07-01

    Full Text Available Objetivo: Determinar la diversidad genética del gen que codifica la proteína rica en glutamato (GLURP de Plasmodium falciparum en pacientes con malaria complicada y no complicada circulante en un área del departamento de Loreto, distrito de Maynas. Materiales y métodos: La diversidad genética fue analizada usando reacción en cadena de la polimerasa (PCR en 30 muestras sanguíneas de pacientes con malaria no complicada (MNC y 46 con malaria grave complicada (MGC. Resultados: Ocho genotipos fueron detectados en pacientes con MNC (Genotipo I,II,III, IV,V, VI,VII y VIII y cuatro genotipos en los pacientes con MGC (Genotipo V,VI,VII,VIII. Asimismo, en 50% de las muestras con MNC fueron detectadas infecciones múltiples, a diferencia de las muestras de MGC en donde no se detectó infecciones múltiples. Conclusión: Existe una diversidad genética en esta región del gen GLURP de P. falciparum, para esa época (marzo 1998 - abril 1999 y esa área del país. En tal sentido, nuestros resultados podrían servir de base para llevar a cabo estudios epidemiológicos posteriores, ya que permitiría conocer la distribución de las cepas circulantes en nuestro país.

  15. Factors Influencing Retention of Gen Y and Non-Gen Y Teachers Working at International Schools in Asia

    Science.gov (United States)

    Fong, Hoi Wah Benny

    2018-01-01

    Quantitative studies on international-school teacher retention are few, especially studies that differentiate between Gen Y and non-Gen Y teachers. This article reports on the findings of a study that examined the relationship of job satisfaction factors to the likelihood of contract renewal by international-school teachers. Results from the study…

  16. Heat and power from MicroGen

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1999-10-01

    This paper reports on the design of a domestic gas-fired cogeneration system developed to replace the central heating boiler. Technical details of the MicroGen demonstration unit are given, and the use of a Linear Free Piston Stirling Engine as the prime mover, and the results of modelling studies of energy demand indicating cost savings compared to conventional boilers are discussed. The enhancement of the benefits of micro-cogeneration through use of thermal and power storage and energy demand management, and the impact of micro-cogeneration on energy use in the home are considered. The UK and European Commission's targets for increased cogeneration capacity are noted.

  17. NextGen Future Safety Assessment Game

    Science.gov (United States)

    Ancel, Ersin; Gheorghe, Adrian; Jones, Sharon Monica

    2011-01-01

    The successful implementation of the next generation infrastructure systems requires solid understanding of their technical, social, political and economic aspects along with their interactions. The lack of historical data that relate to the long-term planning of complex systems introduces unique challenges for decision makers and involved stakeholders which in turn result in unsustainable systems. Also, the need to understand the infrastructure at the societal level and capture the interaction between multiple stakeholders becomes important. This paper proposes a methodology in order to develop a holistic approach aiming to provide an alternative subject-matter expert (SME) elicitation and data collection method for future sociotechnical systems. The methodology is adapted to Next Generation Air Transportation System (NextGen) decision making environment in order to demonstrate the benefits of this holistic approach.

  18. Congenital bilateral neuroblastoma (stage IV-S): case report

    International Nuclear Information System (INIS)

    Lee, Jeong Hee; Lee, Hee Jung; Woo, Seong Ku; Lee, Sang Rak; Kim, Heung Sik

    2002-01-01

    Congenital neonatal neuroblastoma is not uncommon but bilateral adrenal neuroblastoma is rare, accounting for about ten percent of neuroblastomas in children. We report the US the MR findings of a stage IV-S congenital bilateral neuroblastoma occurring in a one-day-old neonate

  19. The GenDev Curriculum Development Workshop.

    Science.gov (United States)

    D'cunha, J

    1997-01-01

    This article describes the second Curriculum Development Workshop held in May 1997 at the Asian Institute of Technology (AIT) in Bangkok, Thailand. The workshop aimed to review critically and restructure the Gender and Development Studies (GenDev) curriculum and to assess AIT's role in training gender experts for the region. Participants included 22 people from 16 countries in Asia, Europe, and the US who were teaching graduate students about gender issues and who were activists with nongovernmental organizations working on gender issues. It was determined that the following were required courses: Culture, Knowledge and Gender Relations; Gender, Technology, and Development; Principles of Gender Research and Methodology in Science and Technology; and Gender Analysis and Field Methods. Other suggested core courses included: Gender and Natural Resource Management; Enterprise Management, Technology, and Gender; Gender and Agrarian Reform; Urbanization: A Gender Perspective; Gender-Responsive Development Planning; and Gender and Economic Change: Past and Present Concerns. Participants distinguished between GenDev courses offered to anyone attending AIT and training courses designed to produce gender experts in the region. The aim of training courses for AIT graduate students was to sensitize potential managers, technologists, and others on gender issues and to create awareness of the importance of including gender perspectives within decision-making, policy formation, and implementation. Training courses to produce gender experts should be directed to those with a prior background in gender studies and include gender analysis in field methods. Participants agreed that there should be an independent and autonomous field of gender and development studies. Participants made six recommendations for such a field of study.

  20. Policy-induced market introduction of Generation IV reactor systems

    International Nuclear Information System (INIS)

    Heek, Aliki Irina van; Roelofs, Ferry

    2011-01-01

    Almost 10 years ago the U.S. Department of Energy (DOE) started the Generation IV Initiative (GenIV) with 9 other national governments with a positive ground attitude towards nuclear energy. Some of these Generation IV systems, like the fast reactors, are nearing the demonstration stage. The question on how their market introduction will be implemented becomes increasingly urgent. One main topic for future reactor technologies is the treatment of radioactive waste products. Technological solutions to this issue are being developed. One possible process is the transformation of long-living radioactive nuclides into short living ones; a process known as transmutation, which can be done in a nuclear reactor only. Various Generation IV reactor concepts are suitable for this process, and of these systems most experience has been gained with the sodium-cooled fast reactor (SFR). However, both these first generation SFR plants and their Generation IV successors are designed as electricity generating plants, and therefore supposed to be commercially viable in the electricity markets. Various studies indicate that the generation costs of a combined LWR-(S)FR nuclear generating park (LWR: light water reactor) will be higher than that of an LWR-only park. To investigate the effects of the deployment of the different reactors and fuel cycles on the waste produced, resources used and costs incurred as a function of time, a dynamic fuel cycle assessment is performed. This study will focus on the waste impact of the introduction of a fraction of fast reactors in the European nuclear reactor park with a cost increase as described in the previous paragraph. The nuclear fuel cycle scenario code DANESS is used for this, as well as the nuclear park model of the EU-27 used for the previous study. (orig.)

  1. Metode Transfer Asam Nukleat sebagai Dasar Terapi Gen

    Directory of Open Access Journals (Sweden)

    Novi Silvia Hardiany

    2017-01-01

    Full Text Available Kemajuan ilmu biologi molekuler memberikan manfaat dalam bidang kedokteran untuk mengembangkanterapi gen. Tujuan terapi gen adalah untuk memperbaiki kerusakan gen atau mengganti gen yang rusakdengan gen yang normal. Pemindahan gen dilakukan dengan teknik transfeksi. Transfeksi merupakanproses pemindahan asam nukleat baik menggunakan vektor virus (transduksi atau menggunakan metodenonviral yaitu zat kimia, lipid dan metode fisik. Vektor virus yang digunakan pada transduksi adalahretrovirus, adenovirus, adeno-associated virus (AAV dan herpes simplex virus (HSV. Keberhasilantransfeksi ditentukan oleh berbagai faktor yang dapat dapat dinilai dengan menggunakan reporter sepertigreen fluorescence protein (GFP. Kata Kunci: terapi gen, transfeksi non viral, transduksi, vektor virus   Methods of Nucleic Acid Transfer as Basic Gene Therapy Abstract The advancement of molecular biology provides benefit in the field of medicine to develop genetherapy. The aim of gene therapy is to repair the genetic damage or to replace damaged gene with thenormal gene. Delivery of gene is carried out by transfection technique, a technique to transfer nucleic acidinto eukaryote cells either using viral vectors (known as transduction, and also using non viral methodsuch as chemical substance, lipid and physical method. Some of the viral vectors used in the transductionare retrovirus, adenovirus, Adeno-associated virus (AAV and Herpes Simplex Virus (HSV. The success oftransfection is determined by various factors which can be assessed using several reporters such as GreenFluorescence Protein (GFP. Key words: gene therapy, non viral transfection, transduction, viral vector. Normal 0 false false false IN X-NONE X-NONE

  2. Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae K., E-mail: taekyu8@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Arcilesi, David J., E-mail: arcilesi.1@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Kim, In H., E-mail: ihkim0730@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Sun, Xiaodong, E-mail: sun.200@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Christensen, Richard N., E-mail: rchristensen@uidaho.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Oh, Chang H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Kim, Eung S., E-mail: kes7741@snu.ac.kr [Idaho National Laboratory, Idaho Falls, ID 83402 (United States)

    2016-04-15

    Highlights: • Uncertainty quantification and benchmark study are performed to validate an ANSYS FLUENT computer model for a depressurization process in a high-temperature gas-cooled reactor. • An ANSYS FLUENT computer model of a 1/8th scaled-down geometry of a VHTR hot exit plenum is presented, which is similar to the experimental test facility that has been constructed at The Ohio State University. • Using the computer model of the scaled-down geometry, the effects of the depressurization process and flow oscillations on the subsequent density-driven stratified flow phenomenology are examined computationally. • The effects of the scaled-down hot exit plenum internal structure temperature on the density-driven stratified flow phenomenology are investigated numerically. - Abstract: An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment. Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study

  3. Glycogen Storage Disease Type IV

    DEFF Research Database (Denmark)

    Bendroth-Asmussen, Lisa; Aksglaede, Lise; Gernow, Anne B

    2016-01-01

    molecular genetic analyses confirmed glycogen storage disease Type IV with the finding of compound heterozygosity for 2 mutations (c.691+2T>C and c.1570C>T, p.R524X) in the GBE1 gene. We conclude that glycogen storage disease Type IV can cause early miscarriage and that diagnosis can initially be made...

  4. Modelo poblacional con algoritmos genéticos

    OpenAIRE

    Veliz Quintero, Eduardo; Rodriguez Ojeda, Luis

    2009-01-01

    Para el desarrollo de este trabajo, “MODELO POBLACIONAL CON ALGORITMOS GENÉTICOS”, he investigado la rama de la inteligencia artificial, como son los algoritmos genéticos. Primero presento en forma general los aspectos que envuelven los algoritmos genéticos, parto de la necesidad de optimizar, así como su historia y posibles aplicaciones y luego he cubierto detalladamente todo lo que pude investigar sobre la teoría de los algoritmos genéticos, sus fundamentos matemáticos, tipos de algoritmos ...

  5. Sobre el significado del descubrimiento del gen FOXP2

    OpenAIRE

    Longa Martínez, Víctor Manuel

    2006-01-01

    El reciente descubrimiento del gen FOXP2 ha ofrecido la primera evidencia clara de la base genética del lenguaje, mostrando una correlación inequívoca desde la perspectiva genética entre una versión mutada de F0XP2 y los trastornos lingüísticos de diferente tipo sufridos por una familia inglesa, conocida como KE. El objetivo central del presente trabajo es discutir diferentes aspectos relacionados con tal descubrimiento; especialmente, la discusión del significado de FOXP2 con ...

  6. About the structure and stability of complex carbonates of thorium (IV), cerium (IV), zirconium (IV), hafnium (IV)

    International Nuclear Information System (INIS)

    Dervin, Jacqueline

    1972-01-01

    This research thesis addressed the study of complex carbonates of cations of metals belonging to the IV A column, i.e. thorium (IV), zirconium (IV), hafnium (IV), and also cerium (IV) and uranium (VI), and more particularly focused on ionic compounds formed in solution, and also on the influence of concentration and nature of cations on stability and nature of the formed solid. The author first presents methods used in this study, discusses their precision and scope of validity. She reports the study of the formation of different complex ions which have been highlighted in solution, and the determination of their formation constants. She reports the preparation and study of the stability domain of solid complexes. The next part reports the use of thermogravimetric analysis, IR spectrometry, and crystallography for the structural study of these compounds

  7. Best-practices guidelines for L2PSA development and applications. Volume 2 - Best practices for the Gen II PWR, Gen II BWR L2PSAs. Extension to Gen III reactors

    International Nuclear Information System (INIS)

    Raimond, E.; Durin, T.; Rahni, N.; Meignen, R.; Cranga, M.; Pichereau, F.; Bentaib, A.; Guigueno, Y.; Loeffler, H.; Mildenberger, O.; Lajtha, G.; Santamaria, C.S.; Dienstbier, J.; Rydl, A.; Holmberg, J.E.; Lindholm, I.; Maennistoe, I.; Pauli, E.M.; Dirksen, G.; Grindon, L.; Peers, K.; Hulqvist, G.; Parozzi, F.; Polidoro, F.; Cazzoli, E.; Vitazkova, J.; Burgazzi, L.; Oury, L.; Ngatchou, C.; Siltanen, S.; Niemela, I.; Routamo, T.; Helstroem, P.; Bassi, C.; Brinkman, H.; Seidel, A.; Schubert, B.; Wohlstein, R.; Guentay, S.; Vincon, L.

    2010-01-01

    The objective of this coordinated action was to develop best practice guidelines for the performance of Level 2 PSA methodologies with a view of harmonisation at EU level and to allow meaningful and practical uncertainty evaluations in a Level 2 PSA. Specific relationships with community in charge of nuclear reactor safety (utilities, safety authorities, vendors, and research or services companies) have been established in order to define the current needs in terms of guidelines for level 2 PSA development and applications. An international workshop was organised in Hamburg, with the support of VATTENFALL, in November 2008. The level 2 PSA experts from the ASAMPSA2 project partners have proposed some guidelines for the development and application of L2PSA based on their experience and on information available from international cooperation (EC Severe Accident network of Excellence - SARNET, IAEA standards, OECD-NEA publications and workshop) or open literature. The number of technical issues addressed in the guideline is very large and all are not covered with the same relevancy in the first version of the guideline. This version is submitted for external review in November 2010 by severe accident experts and PSA, especially, from SARNET and OECD-NEA members. The feedback of the external review will be dis cussed during an international open works hop planned in March 2011 and all outcomes will be taken into consideration in the final version of this guideline (June 2011). The guideline includes 3 volumes: - Volume 1 - General considerations on L2PSA. - Volume 2 - Technical recommendations for Gen II and III reactors. - Volume 3 - Specific considerations for future reactor (Gen IV). The recommendations formulated in the guideline should not be considered as 'mandatory' but should help the L2PSA developers to achieve high quality studies with limited time and resources. It may also help the L2PSA reviewers by positioning one specific study in comparison with some

  8. TrayGen: Arranging objects for exhibition and packaging

    KAUST Repository

    Yang, Yongliang; Huang, Qixing

    2013-01-01

    We present a framework, called TrayGen, to generate tray designs for the exhibition and packaging of a collection of objects. Based on principles from shape perception and visual merchandising, we abstract a number of design guidelines on how

  9. EPCGen2 Pseudorandom Number Generators: Analysis of J3Gen

    Directory of Open Access Journals (Sweden)

    Alberto Peinado

    2014-04-01

    Full Text Available This paper analyzes the cryptographic security of J3Gen, a promising pseudo random number generator for low-cost passive Radio Frequency Identification (RFID tags. Although J3Gen has been shown to fulfill the randomness criteria set by the EPCglobal Gen2 standard and is intended for security applications, we describe here two cryptanalytic attacks that question its security claims: (i a probabilistic attack based on solving linear equation systems; and (ii a deterministic attack based on the decimation of the output sequence. Numerical results, supported by simulations, show that for the specific recommended values of the configurable parameters, a low number of intercepted output bits are enough to break J3Gen. We then make some recommendations that address these issues.

  10. National Project Management Corp (United States). [Technical and regulatory aspects

    International Nuclear Information System (INIS)

    DeBor, Joseph

    2013-01-01

    Studies of the VHTR fuel cycle which involves: (i) use of light water cooled reactor (LWR) spent fuel as kernel feedstock; (ii) recycle of spent VHTR fuel; (iii) use of the VHTR in the management of transuranics (TRU); and, (iv) the geologic storage performance of spent VHTR fuel

  11. Introducing AstroGen: The Astronomy Genealogy Project

    OpenAIRE

    Tenn, Joseph S.

    2016-01-01

    The Astronomy Genealogy Project ("AstroGen"), a project of the Historical Astronomy Division of the American Astronomical Society (AAS), will soon appear on the AAS website. Ultimately, it will list the world's astronomers with their highest degrees, theses for those who wrote them, academic advisors (supervisors), universities, and links to the astronomers or their obituaries, their theses when on-line, and more. At present the AstroGen team is working on those who earned doctorates with ast...

  12. Parapedobacter koreensis gen. nov., sp. nov.

    Science.gov (United States)

    Kim, Myung Kyum; Na, Ju-Ryun; Cho, Dong Ha; Soung, Nak-Kyun; Yang, Deok-Chun

    2007-06-01

    Strain Jip14(T), a Gram-negative, non-spore-forming, rod-shaped, non-motile bacterium, was isolated from dried rice straw and characterized in order to determine its taxonomic position. 16S rRNA gene sequence analysis revealed that strain Jip14(T) belongs to the family Sphingobacteriaceae, and the highest degree of sequence similarity was determined to be to Pedobacter saltans DSM 12145(T) (88.5 %), Pedobacter africanus DSM 12126(T) (87.6 %), Pedobacter heparinus DSM 2366(T) (87.1 %) and Pedobacter caeni LMG 22862(T) (86.9 %). Chemotaxonomic data revealed that strain Jip14(T) possesses menaquinone MK-7 and the predominant fatty acids C(15 : 0) iso, C(16 : 0), C(16 : 0) 10-methyl, C(17 : 0) iso 3-OH and summed feature 3 (C(15 : 0) iso 2-OH/C(16 : 1)omega7c). The results of physiological and biochemical tests clearly demonstrated that strain Jip14(T) represents a distinct species. Based on these data, Jip14(T) should be classified within a novel genus and species, for which the name Parapedobacter koreensis gen. nov., sp. nov. is proposed. The type strain of Parapedobacter koreensis is Jip14(T) (=KCTC 12643(T)=LMG 23493(T)).

  13. Banque Cantonale de Genève

    CERN Multimedia

    Banque Cantonale de Genève

    2011-01-01

    7e Salon Immobilier BCGE le samedi 3 septembre 2011, de 8 h 30 à 13 h 00, au Centre de formation de Conches À cette occasion, les meilleurs spécialistes professionnels genevois de l’immobilier seront réunis en un seul et même lieu. Si vous le souhaitez, un conseiller spécialisé dans les financements hypothécaires évaluera vos possibilités d’investissement immobilier adaptées à votre situation personnelle. En parallèle, les plus importantes régies immobilières de Genève seront à votre disposition pour vous présenter leurs offres actuelles, ainsi que les projets immobiliers futurs et discuter avec vous de la meilleure stratégie à adopter pour trouver l’objet de vos rêves. De plus, vous aurez la possibilité...

  14. Direct Bandgap Group IV Materials

    Science.gov (United States)

    2016-01-21

    AFRL-AFOSR-JP-TR-2017-0049 Direct Bandgap group IV Materials Hung Hsiang Cheng NATIONAL TAIWAN UNIVERSITY Final Report 01/21/2016 DISTRIBUTION A...NAME(S) AND ADDRESS(ES) NATIONAL TAIWAN UNIVERSITY 1 ROOSEVELT RD. SEC. 4 TAIPEI CITY, 10617 TW 8. PERFORMING ORGANIZATION REPORT NUMBER 9. SPONSORING...14. ABSTRACT Direct bandgap group IV materials have been long sought for in both academia and industry for the implementation of photonic devices

  15. Estimación de parámetros genéticos para características productivas y reproductivas en los sistemas doble propósito del trópico bajo colombiano

    Directory of Open Access Journals (Sweden)

    A. P. Galeano

    2010-01-01

    Full Text Available Con el objetivo de estimar los componentes de varianza, las heredabilidades, repetibilidades y correlaciones genéticas y fenotípicas para la producción de leche por lactancia (PL, el peso al destete (PD, el intervalo entre partos (IEP y el Índice de Vaca (IV, de las hembras bovinas manejadas en los sistemas de producción de doble propósito del trópico bajo colombiano, se analizaron los registros productivos y reproductivos de 1.687 vacas registradas en la Asociación Colombiana de Criadores de Ganado en Doble Propósito (Asodoble, durante el periodo comprendido entre 1998 y 2007. Se empleó un modelo animal mixto que incluyó los efectos fijos del grupo contemporáneo (finca-sexo-época-año, la composición racial, y la duración de la lactancia como covariable; así como los efectos genéticos aleatorios del animal, el medio ambiente permanente y el residual. Las heredabilidades estimadas para IEP (0,04 y PD (0,11 fueron bajas, y moderadas para PL (0,35 e IV (0,24, respectivamente. La repetibilidad estimada para IEP fue baja (0,08, y para PL (0,41 e IV (0,31 moderada; en el caso de PD este valor fue igual a la heredabilidad (0,11. Las correlaciones genéticas y fenotípicas obtenidas entre PL y PD con respecto a IEP fueron positivas, y se determinó una asociación genética negativa entre PL y PD. Los resultados demostraron que el IV es un buen indicador, desde el punto de vista genético, de la eficiencia productiva y reproductiva de los animales manejados en estos sistemas productivos.

  16. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  17. Seleção preliminar de genótipos de pinheira em Bom Jesus-PI Preliminary selection of sugar apple genotypes in Bom Jesus county, Piauí state, Brazil

    Directory of Open Access Journals (Sweden)

    Ítalo Herbert Lucena Cavalcante

    2011-01-01

    Full Text Available A pinheira (Annona squamosa L. ocorre espontaneamente no Nordeste Brasileiro, onde é explorada de forma extrativista, caracterizando-se pela falta de manejo adequado e material genético selecionado. Nesse sentido, foi realizado um experimento com objetivo de avaliar a produtividade, as características físicas e químicas de frutos de dez genótipos de pinheira no município de Bom Jesus, PI. Adotou-se delineamento inteiramente casualizado, com tratamentos representados por dez genótipos de pinheira e três repetições. Foram avaliadas as seguintes variáveis: vitamina C, acidez titulável, sólidos solúveis, relação SS/AT "ratio", diâmetros longitudinal e transversal, relação DL/DT, número se sementes por fruto, massa dos frutos e produção por planta. Os genótipos apresentam diferenças quanto às características químicas, físicas e produtivas dos frutos. Os genótipos foram agrupados em sete grupos, com destaque para o grupo III (Gen-02 e grupo IV (Gen-05, fato que explicitou as diferenças entre os genótipos de pinheira quanto às características produtivas e químicas e físicas dos frutos. Genótipos Gen-01 e Gen-02 apresentam potencial para instalação em plantios comerciais, pela produtividade, formato do fruto ou por caracterizarem fontes naturais de vitamina C.The sugar apple (Annona squamosa L. is native to tropical America, occurring spontaneously in Northeastern Brazil, where it is exploited mainly as subsistence without adequate management and without genetic material selection. An experiment was developed aiming to evaluate yield, physical and chemical characteristics of the fruits of ten sugar apple genotypes in Bom Jesus, Piauí State, Brazil. A completely randomized design with treatments represented by ten genotypes and three replications was adopted. The following variables were evaluated: vitamin C, titratable acidity, soluble solids, SS/TA ratio, longitudinal diameter and transverse, LD/TD, number of

  18. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Final Report

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Final report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Mockups applied to design review of AP600/1000, Construction planning for AP 600, and AP 1000 maintenance evaluation. Proof of concept study also performed for GenIV PBMR models

  19. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Final report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Mockups applied to design review of AP600/1000, Construction planning for AP 600, and AP 1000 maintenance evaluation. Proof of concept study also performed for GenIV PBMR models.

  20. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Vasudevan, Vijay; Carroll, Laura; Sham, Sam

    2015-01-01

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  1. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevan, Vijay [Univ. of Cincinnati, OH (United States); Carroll, Laura [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-06

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  2. ANALISIS GEN HAEMAGGLUTININ PADA VIRUS CAMPAK LIAR

    Directory of Open Access Journals (Sweden)

    Subangkit Subangkit

    2015-05-01

    Full Text Available AbstrakPenyakit Campak disebabkan oleh virus campak yang termasuk genus Morbilivirus dan Family Paramyxoviridae. Penyakit campak masih menjadi masalah kesehatan karena masih ditemukan Kejadian Luar Biasa (KLB di Indonesia. Salah satu penyebab terjadinya KLB tersebut diduga sebagaiakibat perbedaan antigenesitas antara strain vaksin yang digunakan dengan strain virus campak liar yang beredar di Indonesia. Penelitian ini bertujuan mendapatkan gambaran tentang karakteristik genetik gen Haemagglutinin virus campak liar yang ada di Indonesia. Spesimen yang digunakan sebanyak 27 isolat virus penyebab KLB dari 17 propinsi selama periode tahun 2003-2010. Isolat virus dilakukan pemeriksaan secara RT-PCR dan sekuensing dengan metode Sanger. Hasil sekuensing dianalisis dengan menggunakan perangkat lunak Bioedit 7.0 dan MEGA 4.0. Hasil penelitian didapatkan perbedaan 10 asam amino antara virus campak strain vaksin CAM-70 dan virus campak liar pada posisi D416N; K424T; V451M; N455T; V466I; I473T; F476L; Y481S atau Y481N; H495N; G505D. Kesimpulan penelitian ini adalah terdapat perbedaan karakteristik genetik antara virus campak liar di Indonesia berbeda dengan strain virus vaksin CAM-70.Kata kunci : Campak, Analisis Molekuler, Hemagglutinin, CD46AbstractMeasles is caused by virus belonging to the genus Morbilivirus and Family Paramyxoviridae. Measles is still a public health problem because outbreak of measles still found in Indonesia. Outbreak is suspected as a result of differences in antigenicity between vaccine strains used with wild-type measles virus strains circulating in Indonesia. This study aims to get genetic characteristics of wild-type measles virus haemagglutinin gene in Indonesia. The specimens were used 27 viral isolates from 17 provinces period 2003-2010. Viral isolates examined by RT-PCR and sequencing with Sanger method. Sequencing analysis were conducted using Bioedit 7.0 and MEGA 4.0 software. The results showed 10 amino acid differences

  3. Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors (Workshop Report)

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, RE

    2004-07-15

    The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods

  4. Free-format RPG IV

    CERN Document Server

    Martin, Jim

    2013-01-01

    This how-to guide offers a concise and thorough introduction to the increased productivity, better readability, and easier program maintenance that comes with the free-format style of programming in RPG IV. Although free-format information is available in IBM manuals, it is not separated from everything else, thereby requiring hours of tedious research to track down the information needed. This book provides everything one needs to know to write RPG IV in the free-format style, and author Jim Martin not only teaches rules and syntax but also explains how this new style of coding has the pot

  5. Genética e hanseníase

    Directory of Open Access Journals (Sweden)

    Bernardo Beiguelman

    Full Text Available As diferentes linhas de pesquisa utilizadas para investigar a importância dos fatores hereditários humanos na determinação da resistência/suscetibilidade à infecção pelo Mycobacterium leprae foram discutidas no presente trabalho. Uma síntese dessas abordagens permitiu analisar os resultados das investigações sobre associação da hanseníase com polimorfismos genéticos, distribuição familial da hanseníase, prevalência da hanseníase e distância genética, concordância da hanseníase em gêmeos e estudos genéticos sobre a reação de Mitsuda.

  6. Estudios sobre plantas andinas,- IV

    Directory of Open Access Journals (Sweden)

    Cuatrecasas José

    1943-04-01

    Full Text Available Caracteres genéricos:-Capitulo radiado, con flores dimorfas, las externas apenas más largas que las interiores. Invólucro cónico, de brácteas pluriseriadas pero poco numerosas, flojas, lineal-lanceoladas, agudas, de consistencia herbácea incluso en la fructificación, más largas que el resto del capitulo. Receptáculo alveolado ,con el margen de las fositas ,escamoso-lacerado. Corolas exteriores femeninas, liguladas, con tubo muy corto y casi rectas, ,con 4 líneas longitudinales y 3 dientes gruesos y callosos, en 2-3 filas, con frecuencia provistas de un apéndice lineal en la garganta, rudimento de labio superior.

  7. Craniostenose em gêmeos: estudo genético

    Directory of Open Access Journals (Sweden)

    Walter Carlos Pereira

    1968-09-01

    Full Text Available É relatada a ocorrência de formas clínicas diversas de craniostenose em gêmeos de sexo diferente. A menina apresentava obliteração completa da sutura coronaria e dos dois terços anteriores da sutura sagital; no menino a sutura sagital era a única afetada. O estudo genético mostrou que a craniostenose independe de aberrações cromossômicas, indicando ser transmitida por gens recessivos raros de natureza autossômica.

  8. Introducing AstroGen: the Astronomy Genealogy Project

    Science.gov (United States)

    Tenn, Joseph S.

    2016-12-01

    The Astronomy Genealogy Project (AstroGen), a project of the Historical Astronomy Division of the American Astronomical Society (AAS), will soon appear on the AAS website. Ultimately, it will list the world's astronomers with their highest degrees, theses for those who wrote them, academic advisors (supervisors), universities, and links to the astronomers or their obituaries, their theses when online, and more. At present the AstroGen team is working on those who earned doctorates with astronomy-related theses. We show what can be learned already, with just ten countries essentially completed.

  9. La genética de las poblaciones centroamericanas

    OpenAIRE

    Barrantes, Ramiro

    2005-01-01

    Las poblaciones centroamericanas no han sido objeto de muchos estudios genéticos con la excepción de análisis esporádicos de la variación entre y dentro de los grupos amerindios y de origen africano ubicados en el área. No obstante, en los últimos 15 años se efectuaron investigaciones sistemáticas en este sentido incluyendo poblaciones mestizas, particularmente las de Costa Rica y Panamá. En los amerindios se efectuaron estudios detallados de su estructura genética y las relaciones filogenéti...

  10. Genes and proteins of Escherichia coli (GenProtEc).

    Science.gov (United States)

    Riley, M; Space, D B

    1996-01-01

    GenProtEc is a database of Escherichia coli genes and their gene products, classified by type of function and physiological role and with citations to the literature for each. Also present are data on sequence similarities among E.coli proteins with PAM values, percent identity of amino acids, length of alignment and percent aligned. The database is available as a PKZip file by ftp from mbl.edu/pub/ecoli.exe. The program runs under MS-DOS on IMB-compatible machines. GenProtEc can also be accessed through the World Wide Web at URL http://mbl.edu/html/ecoli.html.

  11. 11. IV avati Draakoni galeriis...

    Index Scriptorium Estoniae

    2005-01-01

    Tanel Saare (sünd. 1979) näitus "Gott und huhn episode IV: seed shower". Eksponeeritakse väljavõtteid aktsioonidest aastatel 2000-2004 Turus, Nürnbergis, Berliinis, Lohusalus ja Soulis. Osa aktsioone toimus koos rühmitusega Non Grata

  12. Estimación de parámetros genéticos para características productivas y reproductivas en los sistemas doble propósito del trópico bajo colombiano

    Directory of Open Access Journals (Sweden)

    A. P. Galeano

    2010-06-01

    Full Text Available Con el objetivo de estimar los componentes de varianza, las heredabilidades, repetibilidadesy correlaciones genéticas y fenotípicas para la producción de leche por lactancia(PL, el peso al destete (PD, el intervalo entre partos (IEP y el Índice de Vaca (IV,de las hembras bovinas manejadas en los sistemas de producción de doble propósitodel trópico bajo colombiano, se analizaron los registros productivos y reproductivosde 1.687 vacas registradas en la Asociación Colombiana de Criadores de Ganado enDoble Propósito (Asodoble, durante el periodo comprendido entre 1998 y 2007. Seempleó un modelo animal mixto que incluyó los efectos fijos del grupo contemporáneo(finca-sexo-época-año, la composición racial, y la duración de la lactancia comocovariable; así como los efectos genéticos aleatorios del animal, el medio ambientepermanente y el residual. Las heredabilidades estimadas para IEP (0,04 y PD (0,11fueron bajas, y moderadas para PL (0,35 e IV (0,24, respectivamente. La repetibilidadestimada para IEP fue baja (0,08, y para PL (0,41 e IV (0,31 moderada; en el casode PD este valor fue igual a la heredabilidad (0,11. Las correlaciones genéticas y fenotípicasobtenidas entre PL y PD con respecto a IEP fueron positivas, y se determinóuna asociación genética negativa entre PL y PD. Los resultados demostraron que el IVes un buen indicador, desde el punto de vista genético, de la eficiencia productiva yreproductiva de los animales manejados en estos sistemas productivos.

  13. Analysis and design of the SI-simulator software system for the VHTR-SI process by using the object-oriented analysis and object-oriented design methodology

    International Nuclear Information System (INIS)

    Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung

    2008-01-01

    The SI-simulator is an application software system that simulates the dynamic behavior of the VHTR-SI process by the use of mathematical models. Object-oriented analysis (OOA) and object-oriented design (OOD) methodologies were employed for the SI simulator system development. OOA is concerned with developing software engineering requirements and specifications that are expressed as a system's object model (which is composed of a population of interacting objects), as opposed to the traditional data or functional views of systems. OOD techniques are useful for the development of large complex systems. Also, OOA/OOD methodology is usually employed to maximize the reusability and extensibility of a software system. In this paper, we present a design feature for the SI simulator software system by the using methodologies of OOA and OOD

  14. Analisis Mutasi Gen Protein X Virus Hbv Pada Penderita Hepatitis B Akut Di Manado

    OpenAIRE

    Fatimawali; Kepel, Billy

    2014-01-01

    Faktor-faktor yang mempengaruhi perkembangan hepatitis B kronis menjadi kanker hati antara lain mutasi pada gen x. Penelitian ini bertujuan untuk mengidentifikasi gen protein x virus HBV dan menganalisis apakah terjadi mutasi gen yang terkait dengan munculnya tumor ganas sirosis hati (HCC). Penelitian ini menggunakan primer untuk proses nested PCR yang telah dirancang sebelumnya. Proses nested PCR terhadap 10 sampel DNA HBV pasien dilakukan untuk mengamplifikasi fragmen DNA gen x dilanjutkan ...

  15. Fast reactor development and worldwide cooperation in Generation-IV International Forum

    International Nuclear Information System (INIS)

    Sagayama, Yutaka

    2013-01-01

    Objectives of Gen-IV systems development: Goals: Four challenging technology goals have been defined to be applied to innovative nuclear reactor concepts in the 21st century: 1) Safety and Reliability (safe and reliable operation, no offsite emergency response); 2) Sustainability (effective fuel utilization, minimization of nuclear waste); 3) Proliferation Resistance & Physical Protection (to assure unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism); 4) Economic Competitiveness (life-cycle cost advantage over other energy resources). Phase: Each Generation-IV reactor system is one of three stages. 1) Viability Phase; 2) Performance Phase; 3) Demonstration Phase. Target: Commercial Deployment is expected around 2030s or beyond

  16. Epidemiologia genética: epidemiologia, genética ou nenhuma das anteriores?

    Directory of Open Access Journals (Sweden)

    Aguinaldo Gonçalves

    1990-12-01

    Full Text Available No esforço de contribuir para melhor entendimento da identidade da Epidemiologia Genética, são revistas sua concepção, campo de atuação, métodos e técnicas pertinentes e algumas instâncias de aplicação. Entendendo-a como a área de interesse dos fatores genéticos das doenças e suas interações ambientais, apresenta-se seu campo de atuação como constituído por dois segmentos: um descritivo, que lida com conhecimento da distribuição de tais afecções em famílias e populações, seu impacto a nível do coletivo e sua vigilância epidemiológica, bem como o estudo de seus determinantes; o segundo, caracterizado pela intervenção, refere-se às respectivas medidas preventivas. Em que pese possível limitação pela não-consideração de todas as situações existentes, particular atenção é destinada à revisão de métodos e técnicas que possam ser convergentemente aplicados, a partir de procedimentos genéticos e epidemiológicos. Entre eles, destacam-se como estudos de casos tanto metodologias laboratoriais (como os dermatóglifos quanto quantitativos, como cálculo de herdabilidade e análise multivariada. Alguns objetos de estudo são tomados como instância de aplicação, por contarem com investigações específicas em nosso meio: a hanseníase, o hidrargirismo e a esquizofrenia.In an attempt to contribute to a better undestanding of the identity of Genetic Epidemiology, we review its conception, its field of influence, its appropriate methods and techniques and, at last, some of its applications. Genetic Epidemiology involves the study of genetic factors acting on diseases and on their environmental interactions. These includes two major areas: a descriptive one, related to the distribution of such conditions in families and populations, to the epidemiologic surveillance and to the study of determinants; and another characterized by intervention, which is related to preventive measures. Because of the dificulty in

  17. Revision of Corallinaceae (Corallinales, Rhodophyta): recognizing Dawsoniolithon gen. nov., Parvicellularium gen. nov. and Chamberlainoideae subfam. nov. containing Chamberlainium gen. nov. and Pneophyllum.

    Science.gov (United States)

    Caragnano, Annalisa; Foetisch, Alexandra; Maneveldt, Gavin W; Millet, Laurent; Liu, Li-Chia; Lin, Showe-Mei; Rodondi, Graziella; Payri, Claude E

    2018-03-25

    A multi-gene (SSU, LSU, psbA and COI) molecular phylogeny of the family Corallinaceae (excluding the subfamilies Lithophylloideae and Corallinoideae) showed a paraphyletic grouping of six monophyletic clades. Pneophyllum and Spongites were reassessed and recircumscribed using DNA sequence data integrated with morpho-anatomical comparisons of type material and recently collected specimens. We propose Chamberlainoideae subfam. nov., including the type genus Chamberlainium gen. nov., with C. tumidum comb. nov. as the generitype, and Pneophyllum. Chamberlainium is established to include several taxa previously ascribed to Spongites, the generitype of which currently resides in Neogoniolithoideae. Additionally we propose two new genera, Dawsoniolithon gen. nov. (Metagoniolithoideae), with D. conicum comb. nov. as the generitype and Parvicellularium gen. nov. (subfamily incertae sedis), with P. leonardi sp. nov. as the generitype. Chamberlainoideae has no diagnostic morpho-anatomical features that enable one to assign specimens to it without DNA sequence data, and it is the first subfamily to possess both Type 1 (Chamberlainium) and Type 2 (Pneophyllum) tetra/bisporangial conceptacle roof development. Two characters distinguish Chamberlainium from Spongites: tetra/biasporangial conceptacle chamber diameter (300 μm in Spongites) and tetra/bisporangial conceptacle roof thickness (8 cells in Spongites). Two characters also distinguish Pneophyllum from Dawsoniolithon: tetra/bisporangial conceptacle roof thickness (8 cells in Dawsoniolithon) and thallus construction (dimerous in Pneophyllum vs. monomerous in Dawsoniolithon). This article is protected by copyright. All rights reserved. This article is protected by copyright. All rights reserved.

  18. A Novel Role of Human Holliday Junction Resolvase GEN1 in the Maintenance of Centrosome Integrity

    DEFF Research Database (Denmark)

    Gao, M.; Danielsen, Jannie Michaela Rendtlew; Wei, L.-Z.

    2012-01-01

    but not catalytic activity of GEN1 is required for preventing centrosome hyper-amplification, formation of multiple mitotic spindles, and multi-nucleation. Our findings provide novel insight into the biological functions of GEN1 by uncovering an important role of GEN1 in the regulation of centrosome integrity....

  19. Optimal trading strategy for GenCo in LMP-based and bilateral ...

    African Journals Online (AJOL)

    cboonchu

    GenCo) ... In Li and Shahidehpour (2005), a game-based bidding strategy for GenCos with ..... With the different demands, dispatched levels of GenCos vary as shown in Table 6. .... optimisation, AI applications to power systems, and power system ...

  20. GenBank blastn search result: AK064582 [KOME

    Lifescience Database Archive (English)

    Full Text Available AK064582 002-112-F03 AY007820.1 Daucus carota ATPase8 (ATP8) gene, ATP8-Sp1b allele, complete cds; chimeric... ATPase9 (ATP9) gene, ATP9-Sp3 allele, complete cds; and chimeric ATPase6 (ATP6) gen

  1. Safer Systems: A NextGen Aviation Safety Strategic Goal

    Science.gov (United States)

    Darr, Stephen T.; Ricks, Wendell R.; Lemos, Katherine A.

    2008-01-01

    The Joint Planning and Development Office (JPDO), is charged by Congress with developing the concepts and plans for the Next Generation Air Transportation System (NextGen). The National Aviation Safety Strategic Plan (NASSP), developed by the Safety Working Group of the JPDO, focuses on establishing the goals, objectives, and strategies needed to realize the safety objectives of the NextGen Integrated Plan. The three goal areas of the NASSP are Safer Practices, Safer Systems, and Safer Worldwide. Safer Practices emphasizes an integrated, systematic approach to safety risk management through implementation of formalized Safety Management Systems (SMS) that incorporate safety data analysis processes, and the enhancement of methods for ensuring safety is an inherent characteristic of NextGen. Safer Systems emphasizes implementation of safety-enhancing technologies, which will improve safety for human-centered interfaces and enhance the safety of airborne and ground-based systems. Safer Worldwide encourages coordinating the adoption of the safer practices and safer systems technologies, policies and procedures worldwide, such that the maximum level of safety is achieved across air transportation system boundaries. This paper introduces the NASSP and its development, and focuses on the Safer Systems elements of the NASSP, which incorporates three objectives for NextGen systems: 1) provide risk reducing system interfaces, 2) provide safety enhancements for airborne systems, and 3) provide safety enhancements for ground-based systems. The goal of this paper is to expose avionics and air traffic management system developers to NASSP objectives and Safer Systems strategies.

  2. An electronic flight bag for NextGen avionics

    Science.gov (United States)

    Zelazo, D. Eyton

    2012-06-01

    The introduction of the Next Generation Air Transportation System (NextGen) initiative by the Federal Aviation Administration (FAA) will impose new requirements for cockpit avionics. A similar program is also taking place in Europe by the European Organisation for the Safety of Air Navigation (Eurocontrol) called the Single European Sky Air Traffic Management Research (SESAR) initiative. NextGen will require aircraft to utilize Automatic Dependent Surveillance-Broadcast (ADS-B) in/out technology, requiring substantial changes to existing cockpit display systems. There are two ways that aircraft operators can upgrade their aircraft in order to utilize ADS-B technology. The first is to replace existing primary flight displays with new displays that are ADS-B compatible. The second, less costly approach is to install an advanced Class 3 Electronic Flight Bag (EFB) system. The installation of Class 3 EFBs in the cockpit will allow aircraft operators to utilize ADS-B technology in a lesser amount of time with a decreased cost of implementation and will provide additional benefits to the operator. This paper describes a Class 3 EFB, the NexisTM Flight-Intelligence System, which has been designed to allow users a direct interface with NextGen avionics sensors while additionally providing the pilot with all the necessary information to meet NextGen requirements.

  3. Justicia y genética: compensando las diferencias

    Directory of Open Access Journals (Sweden)

    Alejandra Zúñiga-Fajuri

    2013-01-01

    Full Text Available Se analizan los dilemas morales asociados a los avances científicos que en la actualidad nos exigen repensar el concepto de igualdad equitativa de oportunidades. Asimismo, se pasa revista a la discusión filosófica en torno al origen de las desventajas sociales y genéticas que permiten las desigualdades sociales.

  4. Distributed Generation Market Demand Model (dGen): Documentation

    Energy Technology Data Exchange (ETDEWEB)

    Sigrin, Benjamin [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gleason, Michael [National Renewable Energy Lab. (NREL), Golden, CO (United States); Preus, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States); Baring-Gould, Ian [National Renewable Energy Lab. (NREL), Golden, CO (United States); Margolis, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-02-01

    The Distributed Generation Market Demand model (dGen) is a geospatially rich, bottom-up, market-penetration model that simulates the potential adoption of distributed energy resources (DERs) for residential, commercial, and industrial entities in the continental United States through 2050. The National Renewable Energy Laboratory (NREL) developed dGen to analyze the key factors that will affect future market demand for distributed solar, wind, storage, and other DER technologies in the United States. The new model builds off, extends, and replaces NREL's SolarDS model (Denholm et al. 2009a), which simulates the market penetration of distributed PV only. Unlike the SolarDS model, dGen can model various DER technologies under one platform--it currently can simulate the adoption of distributed solar (the dSolar module) and distributed wind (the dWind module) and link with the ReEDS capacity expansion model (Appendix C). The underlying algorithms and datasets in dGen, which improve the representation of customer decision making as well as the spatial resolution of analyses (Figure ES-1), also are improvements over SolarDS.

  5. Meet Mr. and Mrs. Gen X: A New Parent Generation

    Science.gov (United States)

    Howe, Neil

    2010-01-01

    Slowly but surely, Generation Xers have been taking over from Baby Boomers as the majority of parents in elementary and secondary education. In the early 1990s, Gen Xers began joining parent-teacher associations in the nation's elementary schools. Around 2005, they became the majority of middle school parents. By the fall of 2008, they took over…

  6. PowerGen plc report and accounts 1994

    International Nuclear Information System (INIS)

    1994-01-01

    The annual report and accounts of PowerGen plc for the year 1994 are presented. Financial highlights are quoted, followed by the Chairman's statement, reviews by the Chief Executive and Financial Directors, reports by the Auditors and Directors, balance sheets and details of the consolidated profit and loss account and principal accounting policies. A four year summary and shareholder information are included. (UK)

  7. Measuring Gen-Y Customer Experience in the Banking Sector

    Directory of Open Access Journals (Sweden)

    Kyguolienė Asta

    2017-12-01

    Full Text Available The article analyses customer experience as the subject of marketing research and presents methods for assessing customer experience. The results of empirical research revealing the Gen-Y customer experience in using the Lithuanian commercial banks’ services are presented.

  8. A New Parent Generation: Meet Mr. and Mrs. Gen X

    Science.gov (United States)

    Howe, Neil

    2010-01-01

    Slowly but surely, Generation Xers have been taking over from Baby Boomers as the majority of parents in elementary and secondary education. Gen-X parents and Boomer parents belong to two neighboring generations, each possessing its own location in history and its own peer personality. They are similar in some respects, but clearly different in…

  9. GenSVM: a generalized multiclass support vector machine

    NARCIS (Netherlands)

    G.J.J. van den Burg (Gertjan); P.J.F. Groenen (Patrick)

    2016-01-01

    textabstractTraditional extensions of the binary support vector machine (SVM) to multiclass problems are either heuristics or require solving a large dual optimization problem. Here, a generalized multiclass SVM is proposed called GenSVM. In this method classification boundaries for a K-class

  10. Uruguay; 2011 Article IV Consultation

    OpenAIRE

    International Monetary Fund

    2011-01-01

    This 2011 Article IV Consultation highlights that the growth momentum in Uruguay has continued into 2011 but a slowdown is under way, led by weaker exports and slower public investment. Uruguay’s economic and financial vulnerabilities are modest, and the government has reduced debt vulnerabilities significantly and built important financial buffers. Executive Directors have commended authorities’ skillful macroeconomic management that has underpinned Uruguay’s excellent economic performance, ...

  11. Austria; 2013 Article IV Consultation

    OpenAIRE

    International Monetary Fund

    2013-01-01

    This paper presents details of Austria’s 2013 Article IV Consultation. Austria has been growing economically but is facing challenges in the financial sector. Full implementation of medium-term fiscal adjustment plans require specifying several measures and plans that need gradual strengthening to take expected further bank restructuring cost into account. It suggests that strong early bank intervention and resolution tools, a better designed deposit insurance system, and a bank-financed reso...

  12. Pemotongan dan Menyambung DNA dalam Kloning Gen, Studi pada Kloning Gen Prolidase dari Bakteri Asam Laktat

    Directory of Open Access Journals (Sweden)

    Ketut Suriasih

    2015-03-01

    Full Text Available Gene cloning in lactic acid bacteria (LAB is crucial in term to increase their ability to hydrolyze milk protein such as proline. This proline could be hydrolyzed when the LAB undergone cloning on their genome coding the enzyme. The cloning process need technology to separate/isolate the gene capable of proline hydrolyze. Isolation of DNA containing prolidase gene, need DNA genome cutting. After isolation of DNA gene coding prolidase, it is then recombined with other bacterial DNA to obtained recombinant gene. The process need ligase. In gene cloning, knowledge of cutting and joining the DNA should be understood. The enzyme take the role in cutting and joining the DNA were restriction endonuclease and ligase. The restriction enzyme function (1 in inserting a gen into plasmid contained in a vector during gene cloning, and gene expression experiment, and (2 to identify the gene. It is important that the researcher already have standardized  sequenced gene as control. The DNA contained target gene was cut using some restriction enzyme, then the gene was arrayed in electrophoresis gel using southern blot technique. DNA sequence was elucidated by addition of ethydium bromide. To identify/characterize the isolated gene, this DNA sequence was encountered the control DNA.

  13. ASTRID, the SFR Gen IV technology demonstrator project: where are we, where do we stand for? - 15439

    International Nuclear Information System (INIS)

    Rouault, J.; Abonneau, E.; Settimo, D.; Hamy, J.M.; Hayafune, H.; Gefflot, R.; Benard, R.P.; Mandement, O.; Chauveau, T.; Lambert, G.; Audouin, P.; Mochida, H.; Iitsuka, T.; Fukuie, M; Molyneux, J.; Mazel, J.L.

    2015-01-01

    The Preconceptual Design phase (AVP1) of the ASTRID Project ended late 2012, the main goal was to evaluate innovative options. It is now followed by the AVP2 phase planned until the end of 2015 whose objectives are both to focus the design in order to finalize a coherent reactor outline and to finalize by December 2015 the Safety Option Report. The CEA acts as the industrial architect of the project. In 2014, twelve industrial partners were involved in the project. Japan which participates now in the design studies and also in Research/Development in support of the ASTRID Project and VELAN of the French 'Pole Nucleaire de Bourgogne', are the latest partners to join the Project. The Option Selection Process (RCO) is continuing during the AVP2 phase although structuring decisions remain to be made (the choice of the Energy Conversion System between Rankine cycle and Gas Brayton cycle). Other important option selections, which could nevertheless be reconsidered before starting the core of the Basic Design phase are: the choice of an internal fuel storage and a gas fuel handling chain, a rectangular reactor building with a single wall containment, the steam generator size the vertical handling of components. In addition, BOP studies considering the Marcoule site as a possible one are going on. The next important milestone is at the end of 2015 with the release by the Project team of a convincing and coherent Conceptual Design file. (authors)

  14. New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors

    Directory of Open Access Journals (Sweden)

    P. Balestra

    2016-01-01

    Full Text Available The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential, specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead, were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

  15. Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

    2013-06-01

    Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

  16. Identificación de mutaciones puntuales del gen de la 21-hidroxilasa en pacientes afectados con hiperplasia suprarrenal congénita.

    Directory of Open Access Journals (Sweden)

    Dora Fonseca

    2005-06-01

    Full Text Available lntroducción. La hiperplasia suprarrenal congénita es un trastorno autosómico recesivo debido a la inadecuada secreción de cortisol. Mas del 95% de los casos de hiperplasia suprarrenal congénita son causados por defectos del gen de la 21 hidroxilasa, CYP21A2 . Las manifestaciones clínicas incluyen la forma clásica y la forma no clásica. Objetivos. Determinar la frecuencia de las mutaciones puntuales P30L, IVS2-12AIC-G, Del 8pb, I172N, cluster Ex 6, V281L, Q318X, R356W y P453S en pacientes con hiperplasia suprarrenal congénita. Materiales y métodos. Se estudiaron 58 pacientes, de los cuales, 48 fueron clásicos y 10 no clásicos. Mediante PCR alelo-especifica y ACRS (Amplified Creation Restriction Sites, se analizaron 9 mutaciones puntuales del gen CYP21A2 y se determinó la frecuencia en la población analizada. Resultados. Los alelos afectados se identificaron en el 82,8% de los cromosomas. Las mutaciones mas frecuentes fueron: IVS2-12AIC-G (26,7%, Q318X (21,5%, V281L (12,1% e I172N (12,1%. Conclusiones. Las mutaciones mas frecuentes en Colombia son similares a las de otros países del mundo, excepto para Q318X que presentó una mayor frecuencia, pero similar a la de otros países latinoamericanos. Este hallazgo y la existencia de 17,2% de alelos no identificados puede indicar diferencia entre el acervo genético de las poblaciones. En la forma clásica perdedora de sal predominaron las mutaciones Q318X e IVS2-12AIC-G; en la virilizante simple, IVS2-12AIC-G e I172N y en la no clásica , V281L, lo cual esta relacionado con el grado de actividad enzimática. En la forma no clásica, se encontraron alelos severos en el 66,7% de los casos, lo que determina el riesgo de tener hijos afectados con la forma grave virilizante simple o perdedora de sal. Los resultados reportados permiten ofrecer asesoramiento genético y diagnóstico prenatal.

  17. Justicia en salud y genética

    Directory of Open Access Journals (Sweden)

    Maria Graciela De Ortuzar

    2014-06-01

    Full Text Available Las expectativas puestas en el conocimiento genético exceden el ámbito de la medicina tradiciona, debido a que la intervención directa en la lotería natural demandaría el replanteamiento de conceptos centrales de justicia en salud: necesidades médicas, enfermedad, normalidad, e igualdad de oportunidades en el acceso a la salud. El punto en debate es sí el replanteo de dichos conceptos conlleva un cambio radical en las teorías de justicia (libertariana y/o liberal, mostrando su obsolescencia, o sí simplemente se requiere ampliar dichos conceptos claves por fallas estructurales en las mismas teorías. Como hipótesis general considero que los supuestos cuestionamientos, lejos de socavar las bases de las teorías de justicia, sólo ponen en evidencia sus viejos problemas estructurales. Por razones expositivas, dividiré la presentación tres partes. En la Primera parte, analizo la teoría libertariana, estudiando las contradicciones del modelo a través del impacto de la información genética en el seguro privado de salud. En la Segunda Parte, desarrollo la propuesta alternativa liberal rawlsianadanielsiana del modelo de seguro público, evaluando las implicaciones de la genética a partir de la crítica de su concepto biológico de enfermedad y su restricción al acceso a la salud por necesidades naturales. En la Tercera parte presento un modelo integral de necesidades y capacidades básicas, comprendiendo la prevención, el tratamiento y el mejoramiento moralmente permisible (genético y no genético.Mi aporte principal consiste en la elaboración de este modelo normativo integral de necesidades y capacidades para la regulación conjunta de la información y terapia genética con los restantes problemas de salud.

  18. Perda auditiva genética Genetic hearing loss

    Directory of Open Access Journals (Sweden)

    Ricardo Godinho

    2003-01-01

    Full Text Available O progresso das pesquisas relacionadas à perda auditiva genética tem provocado um importante avanço do entendimento dos mecanismos moleculares que governam o desenvolvimento, a função, a resposta ao trauma e o envelhecimento do ouvido interno. Em países desenvolvidos, mais de 50% dos casos de surdez na infância é causada por alterações genéticas e as perdas auditivas relacionadas à idade têm sido associadas com mecanismos genéticos. OBJETIVO: O objetivo desta revisão é relatar as informações mais recentes relacionadas às perdas audtivas de origem genética. FORAMA DE ESTUDO: Revisão sistemática. MATERIAL E MÉTODO: A revisão da literatura inclui artigos indexados à MEDLINE (Biblioteca Nacional de Saúde, NIH-USA e publicados nos últimos 3 anos, além das informações disponíveis na Hereditary Hearing Loss Home Page. CONCLUSÃO: Os recentes avanços no entendimento das perdas auditivas de origem genética têm favorecido a nossa compreensão da função auditiva e tornado o diagnóstico mais apurado. Possivelmente, no futuro, este conhecimento também proporcionará o desenvolvimento de novas terapias para o tratamento das causas genéticas das perdas auditivas.The progress in the research of genetic hearing loss has advanced our understanding of the molecular mechanisms that govern inner ear development, function and response to injury and aging. In the developed world, over 50% of childhood deafness is attributable to genetic causes and even age-related hearing loss has been associated with genetic mechanisms. AIM: The objective of this review is to summarize recent knowledge in genetic hearing loss. STUDY DESIGN: Sistematic review. MATERIAL AND METHODS: The literature review included articles indexed at MEDLINE (The National Library of Medicine, The National Institute of Health - USA focusing on publications from the past 3 years plus the information available at the Hereditary Hearing Loss Home Page. CONCLUSION

  19. Crystalline cerium(IV) phosphates

    International Nuclear Information System (INIS)

    Herman, R.G.; Clearfield, A.

    1976-01-01

    The ion exchange behaviour of seven crystalline cerium(IV) phosphates towards some of the alkali metal cations is described. Only two of the compounds (A and C) possess ion exchange properties in acidic solutions. Four others show some ion exchange characteristics in basic media with some of the alkali cations. Compound G does not behave as an ion exchanger in solutions of pH + , but show very little Na + uptake. Compound E undergoes ion exchange with Na + and Cs + , but not with Li+. Both Li + and Na + are sorbed by compounds A and C. The results are indicative of structures which show steric exclusion phenomena. (author)

  20. PREPARATION OF OXOPORPHINATOMANGANESE (IV) COMPLEX

    Energy Technology Data Exchange (ETDEWEB)

    Willner, I.; Otvos, J.; Calvin, M.

    1980-07-01

    Oxo-manganese-tetraphenylporphyrin (O=Mn{sup IV}-TPP) has been prepared by an oxygen-transfer reaction from iodosylbenzene to MnIITPP and characterized by its i.r. and field desorption mass spectra, which are identical to those of the product obtained by direct oxidation of Mn{sup III}(TPP) in an aqueous medium; it transfers oxygen to triphenylphosphine to produce triphenylphosphine oxide, and it is suggested that similar intermediates are important in oxygen activation by cytochrome P-450 as well as in the photosynthetic evolution of oxygen.

  1. Genética e hanseníase

    OpenAIRE

    Beiguelman Bernardo

    2002-01-01

    As diferentes linhas de pesquisa utilizadas para investigar a importância dos fatores hereditários humanos na determinação da resistência/suscetibilidade à infecção pelo Mycobacterium leprae foram discutidas no presente trabalho. Uma síntese dessas abordagens permitiu analisar os resultados das investigações sobre associação da hanseníase com polimorfismos genéticos, distribuição familial da hanseníase, prevalência da hanseníase e distância genética, concordância da hanseníase em gêmeos e est...

  2. Análise de distância genética entre acessos do gênero Psidium via marcadores ISSR

    Directory of Open Access Journals (Sweden)

    Názila Nayara Silva de Oliveira

    2014-12-01

    Full Text Available O objetivo deste trabalho foi avaliar a distância genética entre 37 acessos da espécie cultivada Psidium guajava, L. (goiaba e de araçás do gênero Psidium do banco de germoplasma da Universidade Estadual do Norte Fluminense (UENF, via marcadores moleculares ISSR. Nos 17 marcadores selecionados, foram obtidas 216 bandas polimórficas. Pelo método de agrupamento UPGMA, houve a formação de cinco principais grupos. Os acessos de araçá da espécie P. cattleyanum Sabine , ficaram alocados nos grupos I e II. No grupo II, foi observada, dentro da espécie P cattleyanum, maior proximidade com a goiabeira. No grupo III, ficou alocado o acesso da espécie P. guineense Sw (araçá-do-campo e dentre os araçás, foi o que ficou mais próximo da goiaba. Os genótipos de goiabeira ficaram alocados do grupo IV e V, confirmando sua alta divergência. Os marcadores moleculares foram eficientes em estimar a distância genética intra e interespecífica.

  3. Estructura y diversidad genética en vacas Holstein de Antioquia usando un polimorfismo del gen bGH

    Directory of Open Access Journals (Sweden)

    Juan Rincon F.

    2013-03-01

    Full Text Available Objetivo. Determinar las frecuencias alélicas y genotípicas del polimorfismo del intrón 3 del gen bGH y estimar algunos parámetros de estructura poblacional en ganado Holstein. Materiales y métodos. El estudio se realizó con 1366 vacas Holstein en 120 hatos de 11 municipios del departamento de Antioquia. Se extrajo DNA por el método de Salting out y la genotipificación se realizó usando la técnica de PCR-RFLPs. La diversidad genética se determinó mediante la comparación de las heterocigosidades, El equilibrio de Hardy-Weinberg (HW y la diferenciación genética entre las poblaciones se realizó usando el software Arlequín 2.0 Las frecuencias alélicas y genotípicas se evaluaron mediante el paquete estadístico SAS®. Resultados. Las frecuencias genotípicas encontradas fueron 0.764 (+/+, 0.223 (+/- y 0.013 (-/- y las frecuencias alélicas 0.876 (+ y 0.124 (-. No se encontraron desviaciones del Equilibrio de Hardy Weinberg en ninguna de las subpoblaciones. La diversidad genética determinada mediante la comparación de las heterocigosidades fue relativamente baja entre poblaciones pero al interior de estas no. El valor de FST de toda la población fue de 0.0068 y significativo (p<0.05, algunos FST pareados también lo fueron, tomando valores desde 0.0 a 0.13. Los estadísticos FIT y FIS no fueron significativos. Conclusiones. El gen bGH es un candidato interesante para evaluar características de importancia económica ya que no parece haber sido sometido a selección directa, presenta una variabilidad media en las poblaciones, observándose diferenciación genética significativa entre distintos municipios, producto de los diferentes sistemas de producción y acceso a las biotecnologías.

  4. Test Review: Advanced Clinical Solutions for WAIS-IV and WMS-IV

    Science.gov (United States)

    Chu, Yiting; Lai, Mark H. C.; Xu, Yining; Zhou, Yuanyuan

    2012-01-01

    The authors review the "Advanced Clinical Solutions for WAIS-IV and WMS-IV". The "Advanced Clinical Solutions (ACS) for the Wechsler Adult Intelligence Scale-Fourth Edition" (WAIS-IV; Wechsler, 2008) and the "Wechsler Memory Scale-Fourth Edition" (WMS-IV; Wechsler, 2009) was published by Pearson in 2009. It is a…

  5. Crusticorallina gen. nov., a nongeniculate genus in the subfamily Corallinoideae (Corallinales, Rhodophyta).

    Science.gov (United States)

    Hind, Katharine R; Gabrielson, Paul W; P Jensen, Cassandra; Martone, Patrick T

    2016-12-01

    Molecular phylogenetic analyses of 18S rDNA (SSU) gene sequences confirm the placement of Crusticorallina gen. nov. in Corallinoideae, the first nongeniculate genus in an otherwise geniculate subfamily. Crusticorallina is distinguished from all other coralline genera by the following suite of morpho-anatomical characters: (i) sunken, uniporate gametangial and bi/tetrasporangial conceptacles, (ii) cells linked by cell fusions, not secondary pit connections, (iii) an epithallus of 1 or 2 cell layers, (iv) a hypothallus that occupies 50% or more of the total thallus thickness, (v) elongate meristematic cells, and (vi) trichocytes absent. Four species are recognized based on rbcL, psbA and COI-5P sequences, C. painei sp. nov., the generitype, C. adhaerens sp. nov., C. nootkana sp. nov. and C. muricata comb. nov., previously known as Pseudolithophyllum muricatum. Type material of Lithophyllum muricatum, basionym of C. muricata, in TRH comprises at least two taxa, and therefore we accept the previously designated lectotype specimen in UC that we sequenced to confirm its identity. Crusticorallina species are very difficult to distinguish using morpho-anatomical and/or habitat characters, although at specific sites, some species may be distinguished by a combination of morpho-anatomy, habitat and biogeography. The Northeast Pacific now boasts six coralline endemic genera, far more than any other region of the world. © 2016 Phycological Society of America.

  6. NextGen Avionics Roadmap Version 2.0

    Science.gov (United States)

    2011-09-30

    Systems Analysis ( IPSA ) Division has defined multiple NextGen Operational (NGOps) Levels, projecting relative performance and risk based on differing...degrees of capability improvements, as shown in Figure 4. IPSA forecasts include the most likely performance NGOps level (i.e., NGOps 3-4), as well...in the near-term. Figures 5 through 9 de- pict the various programs and capabilities aligned with the various NGOps levels. Factors from the IPSA

  7. EVALUATION DE LA SENSIBILITE A Bemisia tabaci (GEN) DE 13 ...

    African Journals Online (AJOL)

    AISA

    champ, le comportement de 13 variétés de tomate contre la pression de Bemisia tabaci (Gen), une mouche vecteur du virus de la jaunisse en cuillère des ... en saison sèche il y a une vingtaine d'années atteignaient 100 % et les pertes de production ... semaines. Le semis s'est effectué sur une planche de 9 m2 (9 x 1 m2).

  8. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2012-01-01

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  9. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

    2012-01-15

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  10. SALOME PLATFORM and TetGen for Polyhedral Mesh Generation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Park, Chan Eok; Kim, Shin Whan [KEPCO E and C Company, Inc., Daejeon (Korea, Republic of)

    2014-05-15

    SPACE and CUPID use the unstructured mesh and they also require reliable mesh generation system. The combination of CAD system and mesh generation system is necessary to cope with a large number of cells and the complex fluid system with structural materials inside. In the past, a CAD system Pro/Engineer and mesh generator Pointwise were evaluated for this application. But, the cost of those commercial CAD and mesh generator is sometimes a great burden. Therefore, efforts have been made to set up a mesh generation system with open source programs. The evaluation of the TetGen has been made in focusing the application for the polyhedral mesh generation. In this paper, SALOME will be evaluated for the efforts in conjunction with TetGen. In section 2, review will be made on the CAD and mesh generation capability of SALOME. SALOME and TetGen codes are being integrated to construct robust polyhedral mesh generator. Edge removal on the flat surface and vertex reattachment to the solid are two challenging tasks. It is worthwhile to point out that the Python script capability of the SALOME should be fully utilized for the future investigation.

  11. The GenABEL Project for statistical genomics.

    Science.gov (United States)

    Karssen, Lennart C; van Duijn, Cornelia M; Aulchenko, Yurii S

    2016-01-01

    Development of free/libre open source software is usually done by a community of people with an interest in the tool. For scientific software, however, this is less often the case. Most scientific software is written by only a few authors, often a student working on a thesis. Once the paper describing the tool has been published, the tool is no longer developed further and is left to its own device. Here we describe the broad, multidisciplinary community we formed around a set of tools for statistical genomics. The GenABEL project for statistical omics actively promotes open interdisciplinary development of statistical methodology and its implementation in efficient and user-friendly software under an open source licence. The software tools developed withing the project collectively make up the GenABEL suite, which currently consists of eleven tools. The open framework of the project actively encourages involvement of the community in all stages, from formulation of methodological ideas to application of software to specific data sets. A web forum is used to channel user questions and discussions, further promoting the use of the GenABEL suite. Developer discussions take place on a dedicated mailing list, and development is further supported by robust development practices including use of public version control, code review and continuous integration. Use of this open science model attracts contributions from users and developers outside the "core team", facilitating agile statistical omics methodology development and fast dissemination.

  12. Variabilidad genética en Prosopis ferox (Mimosaceae

    Directory of Open Access Journals (Sweden)

    Alicia D. Burghardt

    2004-01-01

    Full Text Available Prosopis ferox (Mimosaceae es una especie arbustiva o arbórea espinosa que se distribuye desde el Sur de Bolivia hasta el noroeste de la Argentina. En la provincia de Jujuy se encuentra a grandes alturas (entre los 2400 y los 3700 m s.m.. Existe una gran variabilidad morfológica, especialmente en cuanto a las dimensiones del fruto y la cantidad de semillas por fruto, ambas características importantes debido al uso de esta planta como forraje. Con el objeto de verificar si existe además variabilidad genética, se realizó un estudio electroforético de proteínas seminales de árboles procedentes de distintas localidades de la provincia de Jujuy. Los patrones polipeptídicos obtenidos por SDS-PAGE presentaron en total 26 bandas. Cada población se caracterizó por sus patrones de presencia-ausencia de bandas, habiéndose encontrado variabilidad intrapoblacional (polimorfismo en algunas de ellas, siendo otras genéticamente homogéneas. Los índices polimórficos en poblaciones de P. ferox son comparables a los obtenidos previamente en P. ruscifolia. La variabilidad genética interpoblacional hallada por medio del estudio electroforético de las proteínas seminales hace suponer la existencia de ecotipos

  13. Divergência, variabilidade genética e desempenho agronômico em genótipos de couve.

    OpenAIRE

    Azevedo, Alcinei Mistico

    2012-01-01

    Embora haja grande variabilidade genética para a couve, são poucos trabalhos no Brasil que visão obter informações para programas de melhoramento genético nesta cultura. Assim, objetivou-se neste trabalho caracterizar 30 genótipos de couve a partir de caracteres morfo-agronômicos para estimar a divergência genética, a importância dos caracteres para a divergência, o desempenho agronômico, os parâmetros genéticos e a correlação entre as características avaliadas. O experimento foi conduzido na...

  14. Crescimento de genótipos de frangos tipo caipira

    Directory of Open Access Journals (Sweden)

    R. C. Veloso

    2015-10-01

    Full Text Available RESUMOObjetivou-se com este trabalho comparar o padrão de crescimento, mediante ajustes das respectivas curvas de crescimento por modelos não lineares, bem como estudar o desenvolvimento de cortes de carcaça em relação ao peso da carcaça em diferentes genótipos de frangos tipo caipira. Foram utilizados 840 pintos de um dia, machos, distribuídos em delineamento inteiramente ao acaso, dos seguintes genótipos da linhagem Redbro: Caboclo, Carijó, Colorpak, Gigante Negro, Pesadão Vermelho, Pescoço Pelado e Tricolor. As aves foram alojadas em 28 boxes, sendo 30 aves/boxe, em galpão de alvenaria com acesso a um piquete de 45m², com quatro repetições. O peso corporal individual dos frangos foi medido ao nascer, aos 14, 28, 42, 56, 70 e 84 dias de idade. Para a determinação das curvas de crescimento do peso corporal das aves, os dados coletados foram avaliados por meio dos modelos não lineares: Brody, Gompertz, Logístico, Richards e von Bertalanffy. Foi empregado o PROC NLIN do SAS, utilizando-se o método interativo de Gauss-Newton. Os critérios usados para escolha do modelo de melhor ajuste da curva de crescimento foram o coeficiente de determinação, o desvio padrão assintótico, o desvio médio absoluto dos resíduos e o índice assintótico. As análises para obtenção dos coeficientes alométricos foram realizadas por meio do PROC GLM do SAS para os genótipos Carijó, Colorpak, Pesadão Vermelho, Pescoço Pelado e Tricolor. Foram avaliados os pesos da carcaça, do peito, das coxas, das sobrecoxas, das pernas e das asas das aves abatidas aos 85 dias de idade. Apenas as equações propostas por Gompertz, von Bertalanffy e Logístico atingiram a convergência, e o modelo proposto por von Bertalanffy foi o mais adequado para descrever o crescimento dos genótipos de frangos caipiras. Todos os cortes avaliados apresentaram crescimento tardio em relação ao peso da carcaça em genótipos de frangos tipo caipira.

  15. Some oxozirconium(IV) compounds

    Energy Technology Data Exchange (ETDEWEB)

    Paul, R C; Gupta, S K; Parmar, S S; Vasisht, S K [Punjab Univ., Chandigarh (India). Dept. of Chemistry

    1976-01-01

    Some new oxozirconium(IV) complexes, ZrO(An)/sub 2/, ZrO(Gly)/sub 2/, ZrO(HSal)/sub 2/, ZrO(HPth)/sub 2/, ZrO(Pic)/sub 2/(HPic)/sub 2/, and ZrO(Quin)/sub 2/(HQuin)/sub 2/ have been isolated from the reactions of ZrO(CH/sub 3/COO)/sub 2/CH/sub 3/COOH with anthranilic acid (HAn), glycine (HGly), salicylic acid (H/sub 2/Sal), phthalic acid (H/sub 2/Pth), picolinic acid (HPic), and 8-quinolinol (HQuin) respectively. Their important infrared bands and wherever possible molar conductance and molecular weight have been reported.

  16. J3Gen: A PRNG for Low-Cost Passive RFID

    Directory of Open Access Journals (Sweden)

    Jordi Herrera-Joancomartí

    2013-03-01

    Full Text Available Pseudorandom number generation (PRNG is the main security tool in low-cost passive radio-frequency identification (RFID technologies, such as EPC Gen2. We present a lightweight PRNG design for low-cost passive RFID tags, named J3Gen. J3Gen is based on a linear feedback shift register (LFSR configured with multiple feedback polynomials. The polynomials are alternated during the generation of sequences via a physical source of randomness. J3Gen successfully handles the inherent linearity of LFSR based PRNGs and satisfies the statistical requirements imposed by the EPC Gen2 standard. A hardware implementation of J3Gen is presented and evaluated with regard to different design parameters, defining the key-equivalence security and nonlinearity of the design. The results of a SPICE simulation confirm the power-consumption suitability of the proposal.

  17. Genetics Home Reference: glycogen storage disease type IV

    Science.gov (United States)

    ... Home Health Conditions Glycogen storage disease type IV Glycogen storage disease type IV Printable PDF Open All ... Javascript to view the expand/collapse boxes. Description Glycogen storage disease type IV (GSD IV) is an ...

  18. A cerium(IV)-carbon multiple bond

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Matthew; Lu, Erli; McMaster, Jonathan; Lewis, William; Blake, Alexander J.; Liddle, Stephen T. [Nottingham Univ. (United Kingdom). School of Chemistry

    2013-12-02

    Straightforward access to a cerium(IV)-carbene complex was provided by one-electron oxidation of an anionic ''ate'' cerium(III)-carbene precursor, thereby avoiding decomposition reactions that plague oxidations of neutral cerium(III) compounds. The cerium(IV)-carbene complex is the first lanthanide(IV)-element multiple bond and involves a twofold bonding interaction of two electron pairs between cerium and carbon. [German] Auf direktem Wege zu einem Cer(IV)-Carbenkomplex gelangt man durch die Einelektronenoxidation einer anionischen Carben-Cerat(III)-Vorstufe. So werden Zersetzungsprozesse vermieden, die die Oxidation neutraler Cer(III)-Verbindungen erschweren. Der Cer(IV)-Carbenkomplex enthaelt die erste Lanthanoid(IV)-Element-Mehrfachbindung; dabei binden Cer und Kohlenstoff ueber zwei Elektronenpaare.

  19. iväkoti Riemula

    OpenAIRE

    Alanko, Reetta; Ihanamäki, Katja

    2012-01-01

    Opinnäytetyössä kuvataan yleisesti päivähoidon kehitystä Suomessa sekä päivähoitoa yrittäjän näkökulmasta, tuoden esille sen tämän päivän haasteet ja mahdollisuudet. Työssä on pohdittu yhteistyön merkitystä kunnan kanssa ja sitä, miten kunta voi osaltaan joko rajoittaa tai edesauttaa yksityisen päivähoitoyrityksen toimintaa. Opinnäytetyössä kerrotaan teoriassa Päiväkoti Riemula nimisen, erityispäivähoitopalveluita tarjoavan yrityksen perustamiseen liittyvistä suunnitelmista. Suunnitelluss...

  20. Update History of This Database - GenLibi | LSDB Archive [Life Science Database Archive metadata

    Lifescience Database Archive (English)

    Full Text Available switchLanguage; BLAST Search Image Search Home About Archive Update History Data ...List Contact us GenLibi Update History of This Database Date Update contents 2014/03/25 GenLibi English archi...base Description Download License Update History of This Database Site Policy | Contact Us Update History of This Database - GenLibi | LSDB Archive ... ...ve site is opened. 2007/03/01 GenLibi ( http://gene.biosciencedbc.jp/ ) is opened. About This Database Data

  1. Seleção de genótipos parentais de acerola com base na divergência genética multivariada

    Directory of Open Access Journals (Sweden)

    CARPENTIERI-PÍPOLO VALÉRIA

    2000-01-01

    Full Text Available Este trabalho teve por objetivo identificar e selecionar genótipos parentais de acerola (Malpighia emarginata L. adequadas a programas de melhoramento genético. Nove caracteres quantitativos de maior importância agronômica foram usados para determinação da distância genética e formação de grupos similares de acessos. O agrupamento pelo método de Tocher, a partir das distâncias generalizadas de Mahalanobis, possibilitou a divisão de 14 genótipos em três grupos. Com base na divergência genética e no caráter agronômico-chave (teor de vitamina C, destacaram-se como mais promissores os cruzamentos dos genótipos: AM Mole pertencente ao grupo III, com os genótipos PR AM, N° 18, PR 17, PR 16, Eclipse, AM 22 e Dominga, todos pertencentes ao grupo I.

  2. Extended analysis of Cu IV

    International Nuclear Information System (INIS)

    Meinders, E.; Uijlings, P.

    1980-01-01

    Wavelength data and classifications of 974 Cu IV lines in the region 750-1275 Angstroem are presented. Most of the lines have been classified as transitions from the previously unknown high even configurations 3d 7 5s and 3d 7 4d to 3d 7 4p. The configuration 3d 7 4d is seriously perturbed by 3d 6 4s 2 . The analysis resulted in the identification of 27 levels of 3d 7 5s and 113 levels of (3d 7 4d + 3d 6 4s 2 ) which are reported. The earlier published levels of 3d 7 4s and 3d 7 4p have to be shifted downward as a consequence of improved wavelength data. Radial paramter values, resulting from least-squares fits, are compared to Hartree-Fock values. The eigenvectors obtained in the parametric fitting are used to calculate transition probabilities in intermediate coupling. The relation between the observed intensities of the transitions 3d 7 4d-3d 7 4p and 3d 7 Ss-3d 7 4p is compared to the relation between theoretical values of the transition integrals obtained from Hartree-Fock calculations. A spectroscopic value for the ionization potentials is calculated from the 3d 7 ns configurations. (orig.)

  3. Studies of binary cerium(IV)-praseodymium(IV) and cerium(IV)-terbium(IV) oxides as pigments for ceramic applications

    International Nuclear Information System (INIS)

    Furtado, L.M.L.

    1991-01-01

    It was investigated a series of pigments of general composition Ce 1-x Pr x O 2 , and Ce x Tb y O 2 , exhibiting radish and brown colors, respectively, and high temperature stability. The pigments were obtained by dissolving appropriate amounts of the pure lanthanide oxides in acids and precipitating the rare earths as mixed oxalates, which were isolated and calcined under air, at 1000 0 C. X-Ray powder diffractograms were consistent with a cubic structure for the pigments. Magnetic susceptibility measurements, using Gouy method, indicated the presence of Pr(IV) ions in the Ce 1-x Pr x O 2 pigments and of Terbium predominantly as Tb(III) ions in the Ce-tb mixed oxides. A new method, based on suspension of solid samples in PVA-STB gels (STB = sodium tetradecaborate), was employed for the measurements of the electronic spectra of the pigments. The thermal behaviour the pigments was investigated by the calcination of the oxalates in the temperature range of 500 to 1200 O C, from 10 to 60 minutes. (author)

  4. Distancias genéticas en poblaciones del NOA

    Directory of Open Access Journals (Sweden)

    Acreche, Noemí

    1996-01-01

    Full Text Available La mayor parte de los trabajos realizados en nuestro país sobre polimorfismos hematológicos, abordan la necesaria descripción de las poblaciones. Se pone de relieve la importancia de encarar estudios, en base a la valiosa información publicada, que vinculen los grupos con técnicas que permitan realizar nuevas inferencias sobre sus relaciones. Conocidas en gran medida en cuanto a sus manifestaciones culturales, pueden aportar desde lo genético a la comprensión de los procesos microevolutivos ocurridos en una región. Para el NOA, se ha considerado la presencia de comunidades aborígenes incluídas en cuatro familias lingüísticas. Se tendrán en cuenta estos complejos como representativos de afinidades que se establecen a partir de estrechas relaciones entre las etnias, no sólo por la lengua, sino también por las características de sus sistemas productivos, religiosidad y organización. En base a las frecuencias génicas publicadas correspondientes a los siguientes alelos: I*A, I*B, I*O; M, N, S, s; Dia , Dib; P1, P2; C, c; D, d, E, e; Le, le; Fya, Fyb; Jka, Jkb; K y k se construyeron tablas de frecuencias. Se estimaron los coeficientes de distancias genéticas que fueron analizados y posteriormente incluídos en la construcción de un fenograma de los grupos de estudio, mediante agrupaciones (Sahn Cluster secuenciales, aglomerativas, jerárquicas y anidadas. De acuerdo a la información recopilada de las frecuencias de los 25 alelos estudiados en trece poblaciones de aborígenes del NOA y Paraguay, las distancias genéticas obtenidas reflejan los caracteres lingüístico-culturales.

  5. An Antarctic hypotrichous ciliate, Parasterkiella thompsoni (Foissner) nov. gen., nov. comb., recorded in Argentinean peat-bogs: morphology, morphogenesis, and molecular phylogeny.

    Science.gov (United States)

    Küppers, Gabriela Cristina; Paiva, Thiago da Silva; Borges, Bárbara do Nascimento; Harada, Maria Lúcia; Garraza, Gabriela González; Mataloni, Gabriela

    2011-05-01

    The ciliate Parasterkiella thompsoni (Foissner, 1996) nov. gen., nov. comb. was originally described from Antarctica. In the present study, we report the morphology, morphogenesis during cell division, and molecular phylogeny inferred from the 18S-rDNA sequence of a population isolated from the Rancho Hambre peat bog, Tierra del Fuego Province (Argentina). The study is based on live and protargol-impregnated specimens. Molecular phylogeny was inferred from trees constructed by means of the maximum parsimony, neighbor joining, and Bayesian analyses. The interphase morphology matches the original description of the species. During the cell division, stomatogenesis begins with the de novo proliferation of two fields of basal bodies, each one left of the postoral ventral cirri and of transverse cirri, which later unify. Primordia IV-VI of the proter develop from disaggregation of cirrus IV/3, while primordium IV of the opisthe develops from cirrus IV/2 and primordia V and VI from cirrus V/4. Dorsal morphogenesis occurs in the Urosomoida pattern-that is, the fragmentation of kinety 3 is lacking. Three macronuclear nodules are generated before cytokinesis. Phylogenetic analyses consistently placed P. thompsoni within the stylonychines. New data on the morphogenesis of the dorsal ciliature justifies the transference of Sterkiella thompsoni to a new genus Parasterkiella. Copyright © 2011 Elsevier GmbH. All rights reserved.

  6. La fiction de Madame de Genlis, espace d’interrogation

    Directory of Open Access Journals (Sweden)

    Isabelle Tremblay

    2013-09-01

    Full Text Available The novel constitutes an ideal genre in which to discuss a central question in the eighteenth century: virtue. As opposed to her contemporaries, Mme de Genlis depicts virtue as a state of mind attained through independence and self-confidence rather than through the atonement of one’s faults. How do her heroines realize their potential? Are they able to find a middle ground between their perception of their roles and identity and the expectations that weigh on them? What paths are available to them?

  7. Sistema genérico de replicación

    OpenAIRE

    Jiménez Ortiz, Raúl

    2014-01-01

    Este proyecto se desarrolla en una de las principales compañías eléctricas de España, y trata sobre la creación de un sistema genérico de replicación de datos. La idea surge de la necesidad de estandarizar la forma de traspasar información entre departamentos o sistemas, debido a que existen gran variedad de interfaces diferentes, y cada vez que surge la necesidad de crear una implica nuevos costes de desarrollo, mantenimiento y futuros evolutivos. Entre los diferentes departamentos de ...

  8. TrayGen: Arranging objects for exhibition and packaging

    KAUST Repository

    Yang, Yongliang

    2013-10-01

    We present a framework, called TrayGen, to generate tray designs for the exhibition and packaging of a collection of objects. Based on principles from shape perception and visual merchandising, we abstract a number of design guidelines on how to organize the objects on the tray for the exhibition of their individual features and mutual relationships. Our framework realizes these guidelines by analyzing geometric shapes of the objects and optimizing their arrangement. We demonstrate that the resultant tray designs not only save space, but also highlight the characteristic of each object and the inter-relations between objects. © 2013 The Eurographics Association and John Wiley & Sons Ltd.

  9. Justicia en salud y genética

    OpenAIRE

    Maria Graciela De Ortuzar

    2014-01-01

    Las expectativas puestas en el conocimiento genético exceden el ámbito de la medicina tradiciona, debido a que la intervención directa en la lotería natural demandaría el replanteamiento de conceptos centrales de justicia en salud: necesidades médicas, enfermedad, normalidad, e igualdad de oportunidades en el acceso a la salud. El punto en debate es sí el replanteo de dichos conceptos conlleva un cambio radical en las teorías de justicia (libertariana y/o liberal), mostrando su obsolescencia,...

  10. Zirconium (IV) complexes with some polymethylenediimines | Na ...

    African Journals Online (AJOL)

    The syntheses of zirconium (IV) complexes have been carried out by the reaction of oxozirconium (IV) chloride with the appropriate diimines (Schiff bases). The complexes were isolated as yellow solids which are stable to heat. The complexes were found to be insoluble in most solvents. The infrared spectra, elemental ...

  11. Astragaloside IV liposomes ameliorates adriamycin-induced ...

    African Journals Online (AJOL)

    Methods: The rats were given a single tail intravenous injection of adriamycin (6 mg/kg) within 1 week, and then divided into four groups including normal, model, benazepril and astragaloside IV liposomes group. They were all orally administered dosage of benazepril and astragaloside IV liposomes once daily for 8 weeks.

  12. Thermodynamic data for predicting concentrations of Th(IV), U(IV), Np(IV), and Pu(IV) in geologic environments

    Energy Technology Data Exchange (ETDEWEB)

    Rai, Dhanpat; Roa, Linfeng; Weger, H.T.; Felmy, A.R. [Battelle, Pacific Northwest National Laboratory (PNNL) (United States); Choppin, G.R. [Florida State University (United States); Yui, Mikazu [Waste Isolation Research Division, Tokai Works, Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    1999-01-01

    This report provides thermodynamic data for predicting concentrations of Th(IV), U(IV), Np(IV), and Pu(IV) in geologic environments, and contributes to an integration of the JNC chemical thermodynamic database, JNC-TDB (previously PNC-TDB), for the performance analysis of geological isolation system for high-level radioactive wastes. Thermodynamic data for the formation of complexes or compounds with hydroxide, chloride, fluoride, carbonate, nitrate, sulfate and phosphate are discussed in this report. Where data for specific actinide(IV) species was lacking, the data were selected based on chemical analogy to other tetravalent actinides. In this study, the Pitzer ion-interaction model is used to extrapolate thermodynamic constants to zero ionic strength at 25degC. (author)

  13. Basis for the safety approach for design and assessment of Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Fiorini, G.L.; Leahy, T.

    2009-01-01

    The primary objective of the RSWG is the implementation of a harmonized approach on long-term safety, and to address risk and regulatory issues in development of the next generation of nuclear systems. To this end, the group is proposing safety goals and evaluation methodology applicable for the design and assessment of future systems. The paper resumes the content of the first RSWG report which provides insights for the safety approach and assists the GIF Systems Steering Committee as well as the GIF Experts Group and the GIF Policy Group for the definition of the most adequate safety related Gen IV R and D. The document is also an essential contributor to help identifying the needed supportive crosscut R and D effort (i.e. applicable to all the innovative nuclear technologies). Although the report presents a number of thoughts and recommendations, it really represents only the start of the efforts for the RSWG. (author)

  14. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Application

    International Nuclear Information System (INIS)

    Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Jung, Y. H.; Bang, B. G.

    2006-08-01

    The systematic study was performed to develop the advanced corrosion-resistant Zr alloys for high burnup and Gen IV application. The corrosion behavior was significantly changed with the alloy composition and the corrosion environment. In general, the model alloys with a higher alloying elements showed a higher corrosion resistance. Among the model alloys tested in this study, Zr-10Cr-0.2Fe showed the best corrosion resistance regardless of the corrosion condition. The oxide on the higher corrosion-resistant alloy such as Zr-1.0Cr-0.2Fe consisted of mainly columnar grains, and it have a higher tetragonal phase stability. In comparison with other alloys being considered for the SCWR, the Zr alloys showed a lower corrosion rate than ferritic-martensitic steels. The results of this study imply that, at least from a corrosion standpoint, Zr alloys deserve consideration as potential cladding or structural materials in supercritical water cooled reactors

  15. Enfermedades genéticas del ADN mitocondrial humano

    Directory of Open Access Journals (Sweden)

    Solano Abelardo

    2001-01-01

    Full Text Available Las enfermedades mitocondriales son un grupo de trastornos que están producidos por un fallo en el sistema de fosforilación oxidativa (sistema Oxphos, la ruta final del metabolismo energético mitocondrial, con la consiguiente deficiencia en la biosíntesis del trifosfato de adenosina (ATP, por sus siglas en inglés. Parte de los polipéptidos que componen este sistema están codificados en el ácido desoxirribonucleico (DNA mitocondrial y, en los últimos años, se han descrito mutaciones que se han asociado con síndromes clínicos bien definidos. Las características genéticas del DNA mitocondrial, herencia materna, poliplasmia y segregación mitótica, confieren a estas enfermedades propiedades muy particulares. Las manifestaciones clínicas de estas enfermedades son muy heterogéneas y afectan a distintos órganos y tejidos por lo que su correcto diagnóstico implica la obtención de datos clínicos, morfológicos, bioquímicos y genéticos. El texto completo en inglés de este artículo está disponible en: http://www.insp.mx/salud/index.html

  16. PowerGen plc report and accounts 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    Detailed financial results are presented for the United Kingdom power generation company PowerGen for the year ended 2 April 1995. A review is given of operating and financial performance. Significant reductions in operating costs and improvements in productivity have been achieved. Diversity of fuels and plant portfolio has been enhanced by building 3000 MW of gas fired CCGT plant. Investment in coal import facilities has increased access to international coal markets. Environmental performance has improved with further reductions in SO{sub 2}, NO{sub x} and CO{sub 2} emissions. Overseas projects include construction of 990 MW CCGT power station in Portugal and a contract to build the 1200 MW Paiton 2 coal-fired power station in Indonesia. Investment in lignite mining and power generation assets of MIBRAG in Germany is contributing to profits. PowerGen`s five year contract for coal supply was assigned to RJB Mining (UK) Ltd. during the year. Stocks of coal fell during the year and further reductions are expected during 1995/96.

  17. Divergence and genetic variability among superior rubber tree genotypes Divergência e variabilidade genética de genótipos superiores de seringueira

    Directory of Open Access Journals (Sweden)

    Lígia Regina Lima Gouvêa

    2010-02-01

    Full Text Available The objective of this work was to estimate the genetic variability and divergence among 22 superior rubber tree (Hevea sp. genotypes of the IAC 400 series. Univariate and multivariate analyses were performed using eight quantitative traits (descriptors, including yield. In the univariate analyses, the estimated parameters were: genetic and environmental variances; genetic and environmental coefficients of variation; and the variation index. The Mahalanobis generalized distance, the Tocher agglomerative method and canonical variables were used for the multivariate analyses. In the univariate analyses, variability was verified among the genotypes for all the variables evaluated. The Tocher method grouped the genotypes into 11 clusters of dissimilarity. The first four canonical variables explained 87.93% of the cumulative variation. The highest genetic variability was found in rubber yield-related traits, which contributed the most to the genetic divergence. The most divergent pairs of genotypes are suggested for crossbreeding. The genotypes evaluated are suitable for breeding and may be used to continue the IAC rubber tree breeding program.O objetivo deste trabalho foi estimar a divergência e a variabilidade genética entre 22 genótipos superiores de seringueira (Hevea sp. da série IAC 400. Análises univariadas e multivariadas foram realizadas com oito caracteres quantitativos (descritores, incluindo produtividade. Na análise univariada, os parâmetros estimados foram: variâncias genética e ambiental, coeficientes de variação genética e ambiental, e índice de variação. A distância generalizada de Mahalanobis, o método aglomerativo de Tocher e variáveis canônicas foram utilizados nas análises multivariadas. Nas análises univariadas, verificou-se variabilidade entre os genótipos para todas as variáveis avaliadas. O método de Tocher agrupou os genótipos em 11 grupos de dissimilaridade. As quatro primeiras variáveis can

  18. Gridded precipitation dataset for the Rhine basin made with the genRE interpolation method

    NARCIS (Netherlands)

    Osnabrugge, van B.; Uijlenhoet, R.

    2017-01-01

    A high resolution (1.2x1.2km) gridded precipitation dataset with hourly time step that covers the whole Rhine basin for the period 1997-2015. Made from gauge data with the genRE interpolation scheme. See "genRE: A method to extend gridded precipitation climatology datasets in near real-time for

  19. A 48-plex autosomal SNP GenPlex™ assay for human individualization and relationship testing

    DEFF Research Database (Denmark)

    Tomas Mas, Carmen; Børsting, Claus; Morling, Niels

    2012-01-01

    SNPs are being increasingly used by forensic laboratories. Different platforms have been developed for SNP typing. We describe the GenPlex™ HID system protocol, a new SNP-typing platform developed by Applied Biosystems where 48 of the 52 SNPforID SNPs and amelogenin are included. The GenPlex™ HID...

  20. ANALISIS SEKUEN GEN GLUTATION PEROKSIDASE (GPX1 SEBAGAI DETEKSI STRES OKSIDATIF AKIBAT INFEKSI MYCOBACTERIUM TUBERCULOSIS

    Directory of Open Access Journals (Sweden)

    Ari Yuniastuti

    2013-02-01

    Full Text Available Glutation merupakan antioksidan yang berperan dalam fungsi imun, dan diekspresikan secara genetik oleh urutan gen yang membentuk protein enzim Glutation Peroxidase (GPx1. Bila ekspresi gen berubah maka terjadi perubahan fungsi glutation dan kerentanan terhadap stress oksidatif. Metode yang digunakan adalah Kasus-kontrol. Sampel yang digunakan adalah sampel darah. Kelompok kasus adalah sampel darah pasien tuberkulosis paru sedangkan kelompok kontrol adalah sampel darah orang sehat. Pemeriksaan gen Glutation peroxidase (GPx1 menggunakan metode Polymerase Chain Reaction (PCR untuk melihat pita DNA pada pasien tuberkulosis par serta elektroforesis produk PCR-RFLP gen GPx1 kelompok sampel tuberkulosis. Hasil penelitian menunjukkan bahwa tidak terdapat hubungan yang bermakna antara polimorfisme gen GPx1 (p=0,365 pasein tuberkulois dengan individu sehat, sehingga tidak dapat digunakan sebagai alat deteksi kerentanan terhadap stress oksidatif pada pasien tuberkulosis. Perlu penelitian lanjutan yang menggunakan sampel lebih besar dan populasi etnik yang berbeda.

  1. Sistema inmune y genética: un abordaje diferente a la diversidad de anticuerpos.

    OpenAIRE

    Matta Camacho, Nubia Estela

    2011-01-01

    RESUMEN Es común encontrar en los libros de inmunología o de genética un capítulo con el título de “sistema inmune y genética”, sin embargo su asociación se centra en cómo la generación de anticuerpos rompió el paradigma “un gen, una proteína”, pues en el caso de la producción de anticuerpos, un gen produce millones de proteínas. El sistema inmune tiene muchos vínculos con la genética y la herencia; esta asociación se da porque cualquier sustancia o compuesto que produzca un organi...

  2. Eustochomorpha Girault, Neotriadomerus gen. n., and Proarescon gen. n. (Hymenoptera, Mymaridae, early extant lineages in evolution of the family

    Directory of Open Access Journals (Sweden)

    John T. Huber

    2017-06-01

    Full Text Available Eustochomorpha Girault, with one described species, E. haeckeli Girault, from Australia is redescribed. Neotriadomerus Huber, gen. n., is described, together with seven new species, all from Australia: N. burwelli Huber, sp. n., N. crassus Huber, sp. n., N. darlingi Huber, sp. n., N. gloriosus Huber, sp. n., N. longiovipositor Huber, sp. n., N. longissimus Huber, sp. n. (one of the largest species of Mymaridae, and N. powerae Huber, sp. n. Proarescon Huber, gen. n., is described for P. primitivum (Huber, comb. n., transferred from Borneomymar Huber, and P. similis Huber, sp. n., from Thailand. The previously unknown male of Borneomymar madagascar Huber is described and the genus is redescribed from critical point dried and slide mounted specimens. Triadomerini, stat. n., is proposed to include six genera: Borneomymar, Eustochomorpha and Neotriadomerus, and the Cretaceous Carpenteriana Yoshimoto, Macalpinia Yoshimoto and Triadomerus Yoshimoto. Aresconini is proposed to include five (possibly six genera: Arescon Enock, Kikiki Huber and Beardsley, Proarescon Huber and Tinkerbella Huber and Noyes, and the Cretaceous Myanmymar Huber and, tentatively, also Enneagmus Yoshimoto. The two tribes are proposed as being the earliest lineages in Mymaridae, with Neotriadomerus and Triadomerus being sister genera to the remaining extant and extinct genera, respectively.

  3. Adduct formation in Ce(IV) thenolytrifluoroacetonate

    International Nuclear Information System (INIS)

    Anufrieva, S.I.; Polyakova, G.V.; Snezhko, N.I.; Pechurova, N.I.; Martynenko, L.I.; Spitsyn, V.I.

    1982-01-01

    The literature contains no information on adduct formation in Ce(IV) β-diketonates with additional ligands. Since tetrakis-β-diketonates of Ce(IV) have four six-membered chelate rings, we can suppose that the introduction of an additional monodentate or bidentate ligand into the coordination sphere of Ce(IV) β-diketonates would lead to an increase in the coordination number (CN) of the Ce(IV) to nine or ten. The possibility of realization of such a high CN for Ce(IV) has not been proved; a study of adduct formation by Ce(IV) tetrakis-β-diketonates is thus of theoretical interest. Such an investigation might also be of practical interest, because the introduction of an additional ligand into the coordination sphere of a rare-earth β-diketonate usually increases the solubility of the β-diketonate in nonpolar solvents and increases the volatility of the compound; such a modification of the properties is important for various practical purposes. The aim of our work was to study the possibility of separating solid adducts of Ce(IV) tetrakis-thenoyltrifluoroacetonate with certain oxygen-containing and nitrogen-containing donor monodentate and bidentate ligands, and also to investigate their properties. As the β-diketone we used thenoyltrifluoroacetone (HTTFA), since in a parallel investigation it was found that Ce(TTFA) 4 has a high oxidation-reduction stability

  4. Potencial de vida útil pós-colheita de quatro genótipos de melão tipo Galia Potential of postharvest shelf life of four genotypes of Galia type melons

    Directory of Open Access Journals (Sweden)

    Patrícia Lígia Dantas de Morais

    2004-12-01

    Full Text Available Avaliou-se o potencial de vida útil pós-colheita de melões (Cucumis melo L. tipo Galia (genótipos Primal, Solarking, Total e Vicar. Utilizou-se o delineamento inteiramente casualizado em esquema fatorial 4 (genótipos x 4 (tempos de armazenamento: 0, 3, 6 e 9 dias, com três repetições. Os frutos foram colhidos no estádio de maturação IV (predominantemente amarelo e armazenados à temperatura de 20 ± 1ºC e umidade relativa de 50 ± 2%. O genótipo Solarking apresentou uma firmeza média de polpa superior aos demais, do inicio ao final do período de armazenamento. Em todos os genótipos, os valores de sólidos solúveis no início do armazenamento encontraram-se dentro da faixa aceitável para comercialização no mercado externo, havendo pouca variação com o decorrer do período de armazenamento. A aparência interna limitou o tempo de vida útil pós-colheita do genótipo Total em apenas seis dias. Os genótipos Solarking, Vicar e Primal apresentaram maior potencial na conservação pós-colheita, principalmente o híbrido Solarking, que chegou aos nove dias de armazenamento com boa aparência interna.The postharvest life span of Galia (genotypes Primal, Solarking, Total, and Vicar melons (Cucumis melo L. was evaluated by a four (genotypes x four (storage periods: 0, 3, 6, and 9 days factorial experiment, in a completely randomized design with three replications. The fruits were harvested at maturation stage IV (yellow color predominance, and stored under 20 ± 1ºC and HR 50 ± 2%. Solarking's average firmness was better than that of the other genotypes during the experimental period. At the beginning of the experiment all genotypes had soluble solids contents at the level (eight to ten percent required for exportation, with these levels varying slightly during storage. The internal aspect limited to six days the postharvest life span of genotype Total. Solarking, Vicar, and Primal showed great postharvest conservation potential

  5. Market opportunities: U.S. - PADD IV

    International Nuclear Information System (INIS)

    Garner, R.P.

    1997-01-01

    The current supply and demand balance, the short and long term expectations and marketing opportunities for Canadian crude oil in PADD IV, the Rocky Mountain region in the US, were reviewed. It was suggested that market opportunities in PADD IV are derived from the following four factors: (1) crude oil declines within that area, (2) federal regulations, (3) competitive presence with markets, and (4) population growth. The overall conclusion was that Canadian producers and PADD IV refiners will be looking at an ever-growing relationship based on freight equalized world crude prices. 8 tabs., 5 figs

  6. Diorganotin(IV) Complexes with Methionine Methyl Ester. Equilibria ...

    African Journals Online (AJOL)

    IV) (DBT) and diphenyltin(IV) (DPT) was investigated at 25 °C and 0.1 mol dm–3 ionic strength in water for dimethyltin(IV) and in 50 % dioxane–water mixture for dibutyltin(IV) and diphenyltin(IV). Methionine methyl ester forms1:1 and 1:2 ...

  7. Periodontal Disease Part IV: Periodontal Infections

    OpenAIRE

    Turnbull, Robert S.

    1988-01-01

    In Part IV of this article, the author describes two periodontal infections, acute necrotizing ulcerative gingivitis (trench mouth) and periodontal abscess, both acute painful conditions for which patients may seek advice from their family physician rather than their dentist.

  8. Safety assessment for Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Leahy, T.J.

    2012-01-01

    The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Recent RSWG work has focused on the definition of an integrated safety assessment methodology (ISAM) for evaluating the safety of Generation IV systems. ISAM is an integrated 'tool-kit' consisting of 5 analytical techniques that are available and matched to appropriate stages of Generation IV system concept development: 1) qualitative safety features review - QSR, 2) phenomena identification and ranking table - PIRT, 3) objective provision tree - OPT, 4) deterministic and phenomenological analyses - DPA, and 5) probabilistic safety analysis - PSA. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time

  9. Determination of uranium (IV) by flow voltammetry

    International Nuclear Information System (INIS)

    Ding Anqing

    1987-01-01

    According to the quantitative reaction of U(IV) and Fe(III) in H 2 SO 4 as well as the relation between current and concentration of substance detected, U(IV) has been determined indirectly by measurement of the electrolysis current of residual Fe(III). The columniform electrode used is made of glass carbon particles. At the range of U(IV) from a few micrograms to 40 μg, the linear relation is excellent. The relative standard deviation is within ±4%. The interference of Fe(II), Ti(IV) and U(VI) is negligible but of Ti(III) is serious. This method has been successfully applied in the determination of actual samples (both out line and on line). Main advantages of this procedure are rapid, simple, small amount of sample (only at microgram level) and easy to realize automation, able to use for on line or process analysis

  10. IV&V Project Assessment Process Validation

    Science.gov (United States)

    Driskell, Stephen

    2012-01-01

    The Space Launch System (SLS) will launch NASA's Multi-Purpose Crew Vehicle (MPCV). This launch vehicle will provide American launch capability for human exploration and travelling beyond Earth orbit. SLS is designed to be flexible for crew or cargo missions. The first test flight is scheduled for December 2017. The SLS SRR/SDR provided insight into the project development life cycle. NASA IV&V ran the standard Risk Based Assessment and Portfolio Based Risk Assessment to identify analysis tasking for the SLS program. This presentation examines the SLS System Requirements Review/System Definition Review (SRR/SDR), IV&V findings for IV&V process validation correlation to/from the selected IV&V tasking and capabilities. It also provides a reusable IEEE 1012 scorecard for programmatic completeness across the software development life cycle.

  11. Genetics Home Reference: mucopolysaccharidosis type IV

    Science.gov (United States)

    ... enzymes, GAGs accumulate within cells, specifically inside the lysosomes . Lysosomes are compartments in the cell that break down ... that cause molecules to build up inside the lysosomes are called lysosomal storage disorders. In MPS IV, ...

  12. International Conference on NextGen Electronic Technologies

    CERN Document Server

    Thalmann, Nadia; Bhaaskaran, V

    2017-01-01

    This book is a collection of keynote lectures from international experts presented at International Conference on NextGen Electronic Technologies (ICNETS2-2016). ICNETS2 encompasses six symposia covering all aspects of electronics and communications domains, including relevant nano/micro materials and devices . This volume comprises of recent research in areas like computational signal processing analysis, intelligent embedded systems, nanoelectronic materials and devices, optical and microwave technologies, VLSI design: circuits systems and application, and wireless communication networks, and the internet of things. The contents of this book will be useful to researchers, professionals, and students working in the core areas of electronics and their applications, especially to signal processing, embedded systems, and networking.

  13. Els orígens de la tuberculosi

    OpenAIRE

    Bueno i Torrens, David, 1965-

    2014-01-01

    La tuberculosi és una de les primeres malalties infeccioses de la història de la humanitat. Es dedueix del registre fòssil, per les empremtes que deixa en els ossos d"algunes de les persones que l"han patit. Tradicionalment s"ha assumit que la tuberculosi ve del bestiar boví, que la va transmetre per primer cop a les persones durant el neolític. Però l"anàlisi genètica dels bacteris causants d"aquesta malaltia en diversos indrets del món i de bacteris fòssils trobats en mòmies precolombines h...

  14. Commissioning and Performance Analysis of WhisperGen Stirling Engine

    Science.gov (United States)

    Pradip, Prashant Kaliram

    Stirling engine based cogeneration systems have potential to reduce energy consumption and greenhouse gas emission, due to their high cogeneration efficiency and emission control due to steady external combustion. To date, most studies on this unit have focused on performance based on both experimentation and computer models, and lack experimental data for diversified operating ranges. This thesis starts with the commissioning of a WhisperGen Stirling engine with components and instrumentation to evaluate power and thermal performance of the system. Next, a parametric study on primary engine variables, including air, diesel, and coolant flowrate and temperature were carried out to further understand their effect on engine power and efficiency. Then, this trend was validated with the thermodynamic model developed for the energy analysis of a Stirling cycle. Finally, the energy balance of the Stirling engine was compared without and with heat recovery from the engine block and the combustion chamber exhaust.

  15. Festival du rire de Genève

    CERN Document Server

    Staff Association

    2015-01-01

    Connaissez-vous le Festival du rire de Genève ? La deuxième édition aura lieu du 25 au 28 mars 2015 au Casino-Théâtre à Carouge. Côté programmation, Marc Donnet-Monay ouvre les festivités avant trois autres soirées de folie et d’humour que nous vous laissons le soin de découvrir dans le programme : http://www.rire-geneve.ch/#programme. Réduction de 30% sur l’achat de places pour les membres du personnel du CERN. Pour cela, il suffit de se rendre sur la billetterie en ligne de notre site : www.rire-geneve.ch et d’utiliser le code promotionnel. Contacter le secrétariat de l’Association du personnel (Staff.Association@cern.ch) pour connaitre ce code promotionnel.

  16. How Gen Y and Boomers will reshape your agenda.

    Science.gov (United States)

    Hewlett, Sylvia Ann; Sherbin, Laura; Sumberg, Karen

    2009-01-01

    When it comes to workplace preferences, Generation Y workers closely resemble Baby Boomers. Because these two huge cohorts now coexist in the workforce, their shared values will hold sway in the companies that hire them. The authors, from the Center for Work-Life Policy, conducted two large-scale surveys that reveal those values. Gen Ys and Boomers are eager to contribute to positive social change, and they seek out workplaces where they can do that. They expect flexibility and the option to work remotely, but they also want to connect deeply with colleagues. They believe in employer loyalty but desire to embark on learning odysseys. Innovative firms are responding by crafting reward packages that benefit both generations of workers--and their employers.

  17. The sale of National Power and PowerGen

    International Nuclear Information System (INIS)

    1992-06-01

    In March 1991, the Secretary of State for Energy sold approximately 60 per cent of the shares in National Power and PowerGen who generate most of the electricity produced in England and Wales. The Government's overriding objective was to complete the privatisation of the electricity industry during the lifetime of the Parliament. Within that overriding objective, the Department sought: to maximise net proceeds; to deepen share ownership among individuals; to achieve the overall recognition that the sale of the two companies had been a success; and to achieve a modest premium on the issue price following the state of share dealings. This report sets out the results of a National Audit Office examination of how far the Department achieved their objectives for the sale, and how they controlled its costs. (author)

  18. Genética de las epilepsias Genetics of epilepsy

    Directory of Open Access Journals (Sweden)

    Gustavo A. Charria-Ortiz

    2007-01-01

    Full Text Available En años recientes se ha podido definir con gran exactitud la existencia de alteraciones genéticas específicas en una gran variedad de síndromes epilépticos tradicionales. Es decir, por vez primera se ha podido relacionar de manera contundente y predecible la presencia de alteraciones genómicas y/o proteómicas con síndromes epilépticos antes considerados como "idiopáticos". La gran mayoría de dichos defectos han sido encontrados en genes codificadores para canales iónicos y/o receptores de membrana, lo cual en cierto modo confirma la ya antes postulada relevancia que estas estructuras tienen en la actividad electroquímica espontánea neuronal cuyo desajuste conllevaría a ciertas formas de epilepsia. Esta revisión se centra en los aspectos genéticos y clínicos de dichas condiciones y alteraciones. También se revisarán brevemente los estudios más relevantes de la literatura médica según los cuales -aun a pesar de no haberse definido con la misma exactitud el tipo de anomalías etiológicas- puede tranquilamente inferirse el gran componente genético que parece subyacer a la etiología de las epilepsias. Por ultimo se enfatizará en que a pesar de dichos descubrimientos, su aplicación en la práctica clínica diaria aun es muy limitada, no solo por la relativa rareza de algunos de tales síndromes neurológicos sino también por la poca relevancia que hasta ahora ellos han tenido en el manejo médico rutinario de la mayoría de los pacientes. Las posibilidades inmediatas de tales avances -incluida la farmacogenómica-, así como los posibles conflictos éticos en que se podría incurrir serán también brevemente discutidos.In the last few years, the presence of specific genetic abnormalities leading to some of the classical epileptic syndromes has been clearly elucidated. This means that for the first time, it has become possible to create a strong relationship between the presence of specific genomic and/or proteomic

  19. Big Bang à Genève - French version only

    CERN Multimedia

    2005-01-01

    C'est la dernière conférence du cycle organisé par la section de physique de l'Université de Genève à l'occasion de l'Année internationale de la physique. Pour le bouquet final, la section de physique a choisi le grand boum du Big Bang. Intitulée « Big Bang à Genève », la conférence donnée par Laurent Chevalier de l'institut français CEA Saclay évoquera les expériences qui se préparent au CERN avec le LHC. Leur but est de reproduire et d'analyser les conditions qui prévalaient à l'origine de l'Univers, juste après le Big Bang. L'exposé décrira de façon simple les techniques utilisées pour cette exploration, qui démarrera en 2007. Laurent Chevalier se demandera avec le public quels phénomènes nouveaux les physiciens espèrent découvrir dans ce monde inexploré. Comme les précédentes, la conférence débutera par une démonstration de détection de rayons cosmiques dans l'auditoire et l'utilisation de ces signaux pour créer une « musique cosmique », en collaboration avec le Pr...

  20. Dsm-iv hypochondriasis in primary care

    OpenAIRE

    Escobar, JI; Gara, M; Waitzkin, H; Silver, RC; Holman, A; Compton, W

    1998-01-01

    The object of this study was to assess the prevalence and correlates of the DSM-IV diagnosis of hypochondriasis in a primary care setting. A large sample (N = 1456) of primary care users was given a structured interview to make diagnoses of mood, anxiety, and somatoform disorders and estimate levels of disability. The prevalence of hypochondriasis (DSM-IV) was about 3%. Patients with this disorder had higher levels of medically unexplained symptoms (abridged somatization) and were more impair...

  1. COBRA-IV wire wrap data comparisons

    International Nuclear Information System (INIS)

    Donovan, T.E.; George, T.L.; Wheeler, C.L.

    1979-02-01

    Thermal hydraulic analyses of hexagonally packed wire-wrapped fuel assemblies are complicated by the induced crossflow between adjacent subchannels. The COBRA-IV computer code simultaneously solves the hydrodynamics and thermodynamics of fuel assemblies. The modifications and the results are presented which are predicted by the COBRA-IV calculation. Comparisons are made with data measured in five experimental models of a wire-wrapped fuel assembly

  2. Quantificação da divergência genética entre acessos de goibeira por meio da estratégia Ward-MLM

    Directory of Open Access Journals (Sweden)

    Bianca Machado Campos

    2013-06-01

    Full Text Available O presente trabalho teve como objetivo quantificar a divergência genética entre 138 acessos de goiabeira procedentes do banco de germoplasma da Universidade Estadual do Norte Fluminense Darcy Ribeiro (UENF, com base em descritores morfológicos, agronômicos e físico-químicos, por meio do procedimento Ward - Modified Location Model (MLM. Para tanto, foram avaliados 13 descritores, sendo cinco qualitativos (coloração da polpa, superfície do fruto, formato do fruto ao final do pedúnculo, largura do pescoço e uniformidade da cor da polpa e oito quantitativos (massa média do fruto, diâmetro longitudinal do fruto, diâmetro transversal do fruto, rendimento da polpa, teor de sólidos solúveis totais, acidez do fruto, relação teor de sólidos solúveis totais e acidez do fruto e teor de ácido ascórbico. Detectou-se ampla variabilidade genética pelos dados morfológicos, agronômicos e físico-químicos nos 138 acessos de goiaba. Pelo procedimento da função da verossimilhança, determinou-se oito o número ideal de grupos, com um valor de incremento de 67,51. O grupo III foi considerado o mais distante, enquanto os grupos I, II, IV, V e VI, os mais próximos. O procedimento Ward-MLM é uma ferramenta útil para detectar divergência genética e agrupar os acessos utilizando, simultaneamente, variáveis qualitativas e quantitativas.

  3. Synthesis and characterization of chiral thorium(IV) and uranium(IV) benzamidinate complexes

    Energy Technology Data Exchange (ETDEWEB)

    Schoene, Sebastian; Maerz, Juliane; Kaden, Peter; Patzschke, Michael; Ikeda-Ohno, Atsushi [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Chemistry of the F-Elements

    2017-06-01

    Two chiral benzamidinate complexes of tetravalent actinides (Th(IV) and U(IV)) were synthesized using a salt metathesis reaction of the corresponding actinide(IV) tetrachlorides and the potassium salt of the chiral benzamidine (S,S)-N,N-Bis-(1-phenylethyl)-benzamidine ((S)-HPEBA). The structure of the complexes was determined with single crystal X-ray diffraction. These are the first examples of chiral amidinate complexes of actinides.

  4. Description of Sharon gen. nov. for the Chilean species Asaphes amoenus Philippi, 1861 (Coleoptera: Elateridae

    Directory of Open Access Journals (Sweden)

    Elizabeth T. Arias-Bohart

    2015-10-01

    Full Text Available Sharon gen. nov. is here described to include Asaphes? amoenus Philippi, 1861 comb. nov. from Chile. A redescription of the species is based on the female holotype and material from different geographic locations. Candèze (1891 placed Asaphes amoenus and Parasaphes elegans in the suprageneric group Asaphites. We discuss differences between Sharon gen. nov. and Hemicrepidius Germar, 1839, where Asaphes amoenus was later placed by Blackwelder (1944. Based on morphological characters, Sharon gen. nov. appears to be related to Parasaphes Candèze, 1881, Wynarka Calder, 1986, and Tasmanelater Calder, 1996, all from Australia, suggesting Gondwanan relationships.

  5. Estudio de la variabilidad genética en camélidos bolivianos

    OpenAIRE

    Barreta Pinto, Julia

    2013-01-01

    El estudio de los camélidos sudamericanos es de gran interés en los países andinoscomo Perú, Bolivia, Chile, Argentina, debido a su importante valor económico y suimportancia en el mantenimiento y desarrollo de las poblaciones rurales en dichos países. Dada la falta de estudios genéticos centrados en las poblaciones de camélidos quehabitan en Bolivia, y la necesidad de realizar una valoración de la diversidad genética deestas poblaciones, la presente Tesis doctoral ha abordado el estudio gené...

  6. Consideraciones genéticas sobre las dislipidemias y la aterosclerosis

    OpenAIRE

    Julio César Fernández Travieso

    2008-01-01

    La interacción entre factores genéticos y ambientales explican muchos aspectos de la aterosclerosis y las variaciones genéticas constituyen marcadores de riesgo de la enfermedad coronaria (EC), la cual ocupa el primer lugar entre las causas de morbilidad y mortalidad a nivel mundial. La predisposición familiar a padecer EC, junto al avance vertiginoso en técnicas de análisis de ADN y la disponibilidad de secuencias del genoma humano, han orientado la investigación de alteraciones genéticas re...

  7. Conceptos básicos de programación genética

    Directory of Open Access Journals (Sweden)

    José Jesús Martínez Páez

    2001-04-01

    Full Text Available La Programación Genética, PG, es un retoño de los Algoritmos Genéticos, en la cual los cromosomas que sufren la adaptación son en sí mismos programas de computador. Se usan operadores genéticos  especializados que generalizan la recombinación sexual y la mutación, para los programas de computador estructurados en árbol que están bajo adaptación.

  8. Manipulación genética de seres humanos

    OpenAIRE

    Manuel Santos Alcántara

    2006-01-01

    El gran avance que ha tenido la Genética en los últimos años y, particularmente, aquello relacionado con el desciframiento del genoma humano, ha traído a la discusión pública la posibilidad concreta de manipular genéticamente a los seres humanos. El mejoramiento o perfeccionamiento genético de los seres humanos, denominado eugenesia, actualmente se ha convertido técnicamente en una realidad, motivando una profunda reflexión de tipo ético. La pregunta básica es la siguiente: aquello que es téc...

  9. Polimorfismos del gen ob en bovinos de raza holstein en la Comarca Lagunera, México

    OpenAIRE

    Sarai S. Mendoza-Retana; Miguel A. Gallegos-Robles; Uriel González-Salas; José L. García-Hernández; Manuel Fortis-Hernández; Cirilo Vázquez-Vázquez; Héctor I. Trejo-Escareño

    2017-01-01

    La Comarca Lagunera es la cuenca lechera más importante de México. En la actualidad se están utilizando diversas técnicas que permiten evaluar genéticamente el animal a una edad temprana, permitiendo seleccionar futuros reproductores con características deseables. Entre los genes relacionados con la producción de leche, se encuentran el gen Ob también llamado gen Leptina el cual actúa sobre el sistema nervioso central y tejidos periféricos jugando un papel muy importante ...

  10. Discusión: Explicaciones genéticas y psicológicas de la esquizofrenia.Genética de la esperanza

    Directory of Open Access Journals (Sweden)

    Silvio Bolaños-Salvatierra

    2003-01-01

    Full Text Available En este documento se rebaten críticas hechas por Raventós y Jensen al artículo “Genética y comportamiento”. Cuatro temas fueron seleccionados: 1 se determina que los antipsicóticos aparecieron veinte años después de la concepción hereditaria de la esquizofrenia; 2 se considera que la discusión es altamente pertinente, para nada bizantina o irrelevante, debido que persisten prácticas epistémicas riesgosas en los investigadores genético-conductuales; 3 aunque ninguna conducta humana está exenta de influencia constitucional, el enfoque biologicista se ha propasado al pretender explicar genéticamente casi todo, desconfirmando solapadamente la importancia de la historia personal; y, 4 se plantea que la investigación biológica sobrevalora el peso de las anomalías genéticas frente a la historia social, por lo que solo aparenta cautela. Se propone investigar genéticamente la esperanza con el objetivo de saturar a la humanidad con ese tipo de explicaciones, para alcanzar más rápido una convivencia basada en la tolerancia y el respeto.

  11. The 1st reveal of Gen-V nuclear energy. Prospecting investigation of nuclear power 2050 (A2050) for energy innovation in the nuclear industry

    International Nuclear Information System (INIS)

    Woo, Tae Ho; Lee, Soon Ho

    2012-01-01

    The proposed strategy for the future nuclear energy is analyzed. The conventional nuclear power plants (NPPs) are investigated by the 21 st style interdisciplinary research as the information technology (IT), nanotechnology (NT), and biological technology (BT). New kinds of energy production methods as spherical isotropic power reactor (SIPR) and nano lattice power (NLP) are introduced. In addition, the problems of Gen-IV technologies are challenged to be solved, which is the matters of the mechanical and thermal controls of several coolants cases. The simulation result shows the increasing for the usefulness of the business. The core and vessel are very tractable due to moving core vessel (SIPR). The concept of safety system is changed to be submerged into coolant instead of injection concept (SIPR). The commercial fusion energy is realized for mass energy productions (NLP). Eventually, the safety as well as economical status is increased comparing to previous NPPs. (orig.)

  12. The 1{sup st} reveal of Gen-V nuclear energy. Prospecting investigation of nuclear power 2050 (A2050) for energy innovation in the nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering; Lee, Seok Jong [POSCO Engineering and Construction Co., Ltd., Incheon (Korea, Republic of); Lee, Soon Ho [SK Engineering and Construction Co., Ltd., Seoul (Korea, Republic of)

    2012-11-15

    The proposed strategy for the future nuclear energy is analyzed. The conventional nuclear power plants (NPPs) are investigated by the 21{sup st} style interdisciplinary research as the information technology (IT), nanotechnology (NT), and biological technology (BT). New kinds of energy production methods as spherical isotropic power reactor (SIPR) and nano lattice power (NLP) are introduced. In addition, the problems of Gen-IV technologies are challenged to be solved, which is the matters of the mechanical and thermal controls of several coolants cases. The simulation result shows the increasing for the usefulness of the business. The core and vessel are very tractable due to moving core vessel (SIPR). The concept of safety system is changed to be submerged into coolant instead of injection concept (SIPR). The commercial fusion energy is realized for mass energy productions (NLP). Eventually, the safety as well as economical status is increased comparing to previous NPPs. (orig.)

  13. Solubility study of Tc(IV) oxides

    International Nuclear Information System (INIS)

    Liu, D.J.; Fan, X.H.

    2005-01-01

    The deep geological disposal of the high level radioactive wastes is expected to be a safer disposal method in most countries. The long-lived fission product 99 Tc is present in large quantities in nuclear wastes and its chemical behavior in aqueous solution is of considerable interest. Under oxidizing conditions technetium exists as the anionic species TcO 4 - whereas under the reducing conditions, expected to exist in a deep geological repository, it is generally predicted that technetium will be present as TcO 2 ·nH 2 O. Hence, the mobility of Tc(IV) in reducing groundwater may be limited by the solubility of TcO 2 ·nH 2 O under these conditions. Due to this fact it is important to investigate the solubility of TcO 2 ·nH 2 O. The solubility determines the release of radionuclides from waste form and is used as a source term in radionuclide migration analysis in performance assessment of radioactive waste repository. Technetium oxide was prepared by reduction of a technetate solution with Sn 2 + . The solubility of Tc(IV) oxide has been determined in simulated groundwater and redistilled water under aerobic and anaerobic conditions. The effects of pH and CO 3 2- concentration of solution on solubility of Tc(IV) oxide were studied. The concentration of total technetium and Tc(IV) species in the solutions were periodically determined by separating the oxidized and reduced technetium species using a solvent extraction procedure and counting the beta activity of the 99 Tc with a liquid scintillation counter. The experimental results show that the rate of oxidation of Tc(IV) in simulated groundwater and redistilled water is about (1.49-1.86) x 10 -9 mol/(L·d) under aerobic conditions, but Tc(IV) in simulated groundwater and redistilled water is not oxidized under anaerobic conditions. Under aerobic or anaerobic conditions the solubility of Tc(IV) oxide in simulated groundwater and redistilled water is equal on the whole after centrifugation or ultrafiltration. The

  14. Solubility of Tc(IV) oxides

    International Nuclear Information System (INIS)

    Liu, D.J.; Fan, X.H.

    2005-01-01

    Full text of publication follows: The deep geological disposal of the high level radioactive wastes is expected to be a safer disposal method in most countries. The long-lived fission product 99 Tc is present in large quantities in nuclear wastes and its chemical behavior in aqueous solution is of considerable interest. Under the reducing conditions, expected to exist in a deep geological repository, it is generally predicted that technetium will be present as TcO 2 .nH 2 O. The solubility of Tc(IV) is used as a source term in performance assessment of radioactive waste repository. Technetium oxide was prepared by reduction of a technetate solution with Sn 2+ . The solubility of Tc(IV) oxide has been determined in simulated groundwater and re-distilled water under aerobic and anaerobic conditions. The effects of pH and CO 3 2- concentration of solution on solubility of Tc(IV) oxide were studied. The concentration of total technetium and Tc(IV) species in the solutions were periodically determined by separating the oxidized and reduced technetium species using a solvent extraction procedure and counting the beta activity of the 99 Tc with a liquid scintillation counter. The experimental results show that the rate of oxidation of Tc(IV) in simulated groundwater and re-distilled water is about (1.49∼1.86) x 10 -9 mol/(L.d) under aerobic conditions, but Tc(IV) in simulated groundwater and re-distilled water is not oxidized under anaerobic conditions. Under aerobic or anaerobic conditions the solubility of Tc(IV) oxide in simulated groundwater and re-distilled water is equal on the whole after centrifugation or ultrafiltration. The solubility of Tc(IV) oxide decreases with the increase of pH at pH 10 and is pH independent in the range 2 -8 to 10 -9 mol/L at 2 3 2- concentration. These data could be used to estimate the Tc(IV) solubility for cases where solubility limits transport of technetium in reducing environments of high-level waste repositories. (authors)

  15. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  16. High cost of stage IV pressure ulcers.

    Science.gov (United States)

    Brem, Harold; Maggi, Jason; Nierman, David; Rolnitzky, Linda; Bell, David; Rennert, Robert; Golinko, Michael; Yan, Alan; Lyder, Courtney; Vladeck, Bruce

    2010-10-01

    The aim of this study was to calculate and analyze the cost of treatment for stage IV pressure ulcers. A retrospective chart analysis of patients with stage IV pressure ulcers was conducted. Hospital records and treatment outcomes of these patients were followed up for a maximum of 29 months and analyzed. Costs directly related to the treatment of pressure ulcers and their associated complications were calculated. Nineteen patients with stage IV pressure ulcers (11 hospital-acquired and 8 community-acquired) were identified and their charts were reviewed. The average hospital treatment cost associated with stage IV pressure ulcers and related complications was $129,248 for hospital-acquired ulcers during 1 admission, and $124,327 for community-acquired ulcers over an average of 4 admissions. The costs incurred from stage IV pressure ulcers are much greater than previously estimated. Halting the progression of early stage pressure ulcers has the potential to eradicate enormous pain and suffering, save thousands of lives, and reduce health care expenditures by millions of dollars. Copyright © 2010 Elsevier Inc. All rights reserved.

  17. Direct complexonometric determination of thorium (IV), uranium (IV), neptunium (IV), plutonium (IV) by titration of diethylenetriaminepentaacetic acid with xylenol orange as indicator

    International Nuclear Information System (INIS)

    Rykov, A.G.; Piskunov, E.M.; Timofeev, G.A.

    1975-01-01

    The purpose of the present work was to develop a method of determining Th(IV), U(IV), Np(N) and Pu(IV) in acid solutions by titration with diethylenetriamine pentacetic acid, the indicator being xylenol orange. It has been established that Th, U, Np and Pu can be determined to within 0.5-1.5%. Th and U in quantities of tens of milligrams can be determined with greater accuracy, attaining hundredths of one per cent. During titration the determination is not hindered by singly- and doubly-charged metal ions, trivalent lanthanides and actinides, except plutonium. The proposed method can be used to determine U(IV) in the presence of considerable quantities of U(VI) and Np(IV) in the presence of Np(V). Total concentrations of uranium or neptunium are determined by reducing uranium (VI) or neptunium (V) by a standard method (for example, using metallic lead, cadmium or zinc amalgam) to the tetravalent state and applying the method described in the paper. (E.P.)

  18. The Contribution of IVS to IGGOS

    Science.gov (United States)

    Nothnagel, A.

    2002-05-01

    Since its inauguration in 1999, the International VLBI Service for Geodesy and Astrometry has made significant progress in the coordination and utilisation of worldwide VLBI resources. Improving the visibility of the IVS components to a wider public in turn led to a higher motivation of the individuals to contribute to the global effort. Not only the number of IVS components but also their investments in terms of funds and manpower demonstrate the increased awareness of the importance of this joint international endeavour. The different demands of the users but also of the contributors often require the definition of priorities which are only being acceptable due to the existence of a strong umbrella organisation like the IVS. Significant progress has also been made in the area of routine data analysis and combination of results. By now, six IVS Analysis Centers provide the redundancy necessary for a robust combination of the results. The use of ITRF2000 station coordinates as the basis for the IVS combined EOP series is the most recent step towards the generation of a consistent chain from the quasi-inertial frame of radio sources to system Earth.

  19. Solubility studies of Np(IV)

    International Nuclear Information System (INIS)

    Zhang Yingjie; Yao Jun; Jiao Haiyang; Ren Lihong; Zhou Duo; Fan Xianhua

    2001-01-01

    The solubility of Np(IV) in simulated underground water and redistilled water has been measured with the variations of pH(6-12) and storage time (0-100 d) in the presence of reductant (Na 2 S 2 O 4 , metallic Fe). All experiments are performed in a low oxygen concentration glove box containing high purity Ar(99.99%), with an oxygen content of less than 5 x 10 -6 mol/mol. Experimental results show that the variation of pH in solution has little effect on the solubility of Np(IV) in the two kinds of water; the measured solubility of Np(IV) is affected by the composition of solution; with Na 2 S 2 O 4 as a reductant, the solubility of Np(IV) in simulated underground water is (9.23 +- 0.48) x 10 -10 mol/L, and that in redistilled water is (8.31 +- 0.35) x 10 -10 mol/L; with metallic Fe as a reductant, the solubility of Np(IV) in simulated underground water is (1.85 +- 0.56) x 10 -9 mol/L, and that in redistilled water is (1.48 +- 0.66) x 10 -9 mol/L

  20. Characterization of Romboutsia ilealis gen. nov., sp. nov., isolated from the gastro-intestinal tract of a rat, and proposal for the reclassification of five closely related members of the genus Clostridium into the genera Romboutsia gen. nov., Intestinibacter gen. nov., Terrisporobacter gen. nov. and Asaccharospora gen. nov.

    Science.gov (United States)

    Gerritsen, Jacoline; Fuentes, Susana; Grievink, Wieke; van Niftrik, Laura; Tindall, Brian J; Timmerman, Harro M; Rijkers, Ger T; Smidt, Hauke

    2014-05-01

    A Gram-positive staining, rod-shaped, non-motile, spore-forming obligately anaerobic bacterium, designated CRIBT, was isolated from the gastro-intestinal tract of a rat and characterized. The major cellular fatty acids of strain CRIBT were saturated and unsaturated straight-chain C12-C19 fatty acids, with C16:0 being the predominant fatty acid. The polar lipid profile comprised six glycolipids, four phospholipids and one lipid that did not stain with any of the specific spray reagents used. The only quinone was MK-6. The predominating cell-wall sugars were glucose and galactose. The peptidoglycan type of strain CRIBT was A1σ lanthionine-direct. The genomic DNA G+C content of strain CRIBT was 28.1 mol%. On the basis of 16S rRNA gene sequence similarity, strain CRIBT was most closely related to a number of species of the genus Clostridium, including Clostridium lituseburense (97.2%), Clostridium glycolicum (96.2%), Clostridium mayombei (96.2%), Clostridium bartlettii (96.0%) and Clostridium irregulare (95.5%). All these species show very low 16S rRNA gene sequence similarity (genus Clostridium. DNA-DNA hybridization with closely related reference strains indicated reassociation values below 32%. On the basis of phenotypic and genetic studies, a novel genus, Romboutsia gen. nov., is proposed. The novel isolate CRIBT (=DSM 25109T=NIZO 4048T) is proposed as the type strain of the type species, Romboutsia ilealis gen. nov., sp. nov., of the proposed novel genus. It is proposed that C. lituseburense is transferred to this genus as Romboutsia lituseburensis comb. nov. Furthermore, the reclassification into novel genera is proposed for C. bartlettii, as Intestinibacter bartlettii gen. nov., comb. nov. (type species of the genus), C. glycolicum, as Terrisporobacter glycolicus gen. nov., comb. nov. (type species of the genus), C. mayombei, as Terrisporobacter mayombei gen. nov., comb. nov., and C. irregulare, as Asaccharospora irregularis gen. nov., comb. nov. (type species

  1. On the stabilization of niobium(V) solutions by zirconium(IV) and hafnium(IV)

    DEFF Research Database (Denmark)

    Sørensen, E.; Bjerre, A.B.

    1992-01-01

    Niobium cannot be separated from zirconium or hafnium when these elements occur together in solution with common anions such as chloride and sulphate. This is ascribed to the co-polymerization of niobium(V) and the hydrolysed ionic species of zirconium(IV) and hafnium(IV) to form colloidal...

  2. Complexation of the An(IV) by NTA; Complexation des An(IV) par le NTA

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, L. [Paris-11 Univ., 91 - Orsay (France)]|[CEA Valrho, Lab. de Chimie des Actinides (LCA), 30 - Marcoule (France)

    2006-07-01

    In the framework of the Nuclear and Environmental Toxicology program, developed in France, it has been decided to take again the studies concerning the actinides decorporation. A similar study of the neptunium complexation by the citrate ions has been carried out on the complexation of Np(IV) with the nitrilotriacetic acid (NTA). The NTA can be considered as a model molecule of the de-corporating molecules (amino-carboxy- ligand). The results of the spectrophotometric measurements being encouraging, the behaviour of several actinides at the same oxidation state (+IV) (Th(IV), U(IV), Np(IV), and Pu(IV)) has been determined. The experimental results are presented. In order to determine the structure of the complexes of stoichiometry 1:2 An(IV)-(NTA){sub 2} in solution, quantic chemistry calculations and EXAFS measurements have been carried out in parallel. These studies confirm the presence of An(IV)-nitrogen bonds whose length decreases from thorium to plutonium and indicate the presence of a water molecule bound to the thorium and the uranium (coordination number 8 for Np/Pu, 9 for Th/U). The evolution of the complexation constants determined in this study in terms of 1/r (r ionic radius of the cation taking into account its coordination number 8 or 9) confirms the change of the coordination number between Th/U and Np/Pu. (O.M.)

  3. Transuranium perrhenates: Np(IV), Pu(IV) and (III), Am (III)

    International Nuclear Information System (INIS)

    Silvestre, Jean-Paul; Freundlich, William; Pages, Monique

    1977-01-01

    Synthesis in aqueous solution and by solid state reactions, crystallographical characterization and study of the stability of some transuranium perrhenates: Asup(n+)(ReO 4 - )sub(n) (A=Np(IV), Pu(IV), Pu(III), Am(III) [fr

  4. Siamia luxuriosa gen. et sp. nov., a new synnematous hyphomycete from Thailand

    NARCIS (Netherlands)

    Robert, Vincent; Decock, Cony; Castañeda Ruíz, Rafael F.

    2000-01-01

    A new synnematous hyphomycete, Siamia luxuriosa gen. et sp. nov., is described from Khao Yai National Park, Thailand. The genus is characterised by having long, dark synnemata, conidiogenous cells with numerous protuberant, cylindrical, thickened unilaterally arranged scars, and long, slightly

  5. Download - GenLibi | LSDB Archive [Life Science Database Archive metadata

    Lifescience Database Archive (English)

    Full Text Available switchLanguage; BLAST Search Image Search Home About Archive Update History Data ...access [here]. About This Database Database Description Download License Update History of This Database Site Policy | Contact Us Download - GenLibi | LSDB Archive ...

  6. License - GenLibi | LSDB Archive [Life Science Database Archive metadata

    Lifescience Database Archive (English)

    Full Text Available switchLanguage; BLAST Search Image Search Home About Archive Update History Data ...nse Update History of This Database Site Policy | Contact Us License - GenLibi | LSDB Archive ...

  7. Transformação genética em espécies florestais.

    OpenAIRE

    Claudia Studart-Guimarães; Cristiano Lacorte; Ana Cristina Miranda Brasileiro

    2010-01-01

    A transformação genética, que compreende a introdução de genes exógenos de forma controlada no genoma de uma célula vegetal e posterior regeneração da planta transgênica, tem contribuído com os programas de melhoramento genético de plantas pela obtenção de genótipos com novas características de interesse. O melhoramento de espécies florestais é limitado por características intrínsecas a tais espécies, como a altura dos indivíduos e o ciclo longo de vida. A transformação genética constitui, po...

  8. A Software-Assurance Design Approach for NextGen Enabling Technologies, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The Next Generation Air Transportation System (NextGen) brings significant advancements to the current management of the National Airspace (NAS). These fundamental...

  9. Ekspresi Gen CYP19 Aromatase, Estrogen, Androgen pada penderita Periodontitis Agresif

    Directory of Open Access Journals (Sweden)

    Dahlia Herawati

    2016-11-01

    Full Text Available Kepadatan tulang tubuh ditentukan oleh gen CYP19 aromatase, hormon estrogen dan androgen. Pada periodontitis agresif terjadi perkembangan cepat kerusakan tulang alveolar, dan kerusakan tulang alveoler tersebut tidak diimbangioleh regenerasi tulang. Tujuan penelitian ini adalah menunjukkan ekspresi gen CYP19 aromatase, estrogen, androgen pada penderita periodontitis agresif agar dapat untuk menjadi pertimbangan pada saat melakukan perawatan periodontal. Metode penelitian, pemeriksaan ekspresi gen aromatse CYP19 berasal dari spesimen tulang alveolar menggunakan imunohistokimia, pengukuran hormon estrogen dan androgen dari serum menggunakan Vidas: Elfa. Hasil penelitian ekspresi gene CYP19 aromatase pada periodontitis agresif menunjukkan gambaran lebih rendah densitasnya dibandingkan pada nonperiodontitis. Estrogen dan androgen pad aperiodontitis agresif ada kecenderungan lebih rendah dibandingkan pada nonperiodontitis. Kesimpulan regenerasi tulang alveoler pad a periodontitis agresif terhambat karena sedikitnya gen CYP19 aromatase dan hormon estrogen dan androgen yang berperan pada pembentukan tulang alveoler kurang memadai.

  10. Neofomitella polyzonata gen. et sp. nov., and N. fumosipora and N. rhodophaea transferred from Fomitella

    Czech Academy of Sciences Publication Activity Database

    Li, H.J.; Li, X.C.; Vlasák, Josef; Dai, Y.C.

    2014-01-01

    Roč. 129, č. 1 (2014), s. 7-20 ISSN 0093-4666 Institutional support: RVO:60077344 Keywords : Basidiomycota * phylogeny * polyporaceae * taxonomy Subject RIV: EB - Gen etics ; Molecular Biology Impact factor: 0.705, year: 2014

  11. CLONACIÓN Y FILOGENIA MOLECULAR DE UN SEGMENTO DEL GEN CODANTE DE LA ACTINA DE MYRCIARIA DUBIA “CAMU-CAMU”: UN CANDIDATO PARA GEN DE REFERENCIA

    Directory of Open Access Journals (Sweden)

    Juan Carlos Castro Gómez

    2012-12-01

    Full Text Available Myrciaria dubia “camu-camu” es un frutal amazónico caracterizado por su amplia variación de vitamina C. Pero los estudios genético moleculares que puedan explicar esta variación son limitados. Por ello nuestro objetivo fue realizar la clonación y filogenia molecular de un segmento del gen codante de la actina de M. dubia. Las muestras fueron obtenidas de la colección de germoplasma del INIA. Luego, el ARN fue purificado y mediante RT-PCR con cebadores degenerados se amplificó un segmento del gen. En base a la secuencia obtenida se diseñaron cebadores específicos para PCR en tiempo real. Los resultados muestran que se ha aislado, clonado y secuenciado un segmento del gen codante de actina de M. dubia y detectado su expresión en hojas, pulpa y cáscara de M. dubia. Así, con el soporte de herramientas bioinformáticas y uso de técnicas de biología molecular hemos aislado, clonado y secuenciado un segmento del gen codante de la actina de M. dubia. Asimismo, los análisis realizados muestran que el gen se expresa y presenta niveles similares de expresión en hojas, pulpa y cáscara de M. dubia. Sin embargo, es necesario realizar más experimentos a fin de verificar su estabilidad de expresión.

  12. TRANSFER GEN ANTIVIRUS PADA EMBRIO UDANG WINDU, Penaeus monodon DALAM BERBAGAI KONSENTRASI DEOXYRIBO NUCLEIC ACID

    Directory of Open Access Journals (Sweden)

    Andi Parenrengi

    2011-12-01

    Full Text Available Teknologi transgenesis khususnya rekayasa genetik untuk menghasilkan udang windu resisten penyakit merupakan salah satu strategi yang dapat dilakukan dalam upaya pemecahan masalah penyakit yang menimpa budidaya udang windu. Teknologi transgenesis khususnya transfer gen antivirus pada udang windu telah berhasil dilakukan melalui teknik transfeksi. Meskipun demikian optimalisasi komponen teknologi tersebut masih perlu dilakukan. Konsentrasi DNA gen merupakan salah satu komponen teknologi transgenesis yang harus dioptimalkan untuk mendapatkan efisiensi dalam transfer gen. Penelitian bertujuan untuk mengetahui konsentrasi DNA gen antivirus yang optimal sebagai bahan transfer gen ke embrio menggunakan metode transfeksi. Embrio udang windu yang diperoleh dari hasil pemijahan induk asal Aceh, dikoleksi 5-10 menit setelah memijah dengan kepadatan 625 telur/2 mL. Transfeksi dilakukan dengan menggunakan media larutan transfeksi jetPEI dengan konsentrasi DNA gen antivirus sebagai perlakuan, yakni: 5, 10, dan 15 µg serta kontrol positif (tanpa plasmid DNA dan negatif (tanpa plasmid DNA dan larutan transfeksi, masing-masing 3 ulangan. Embrio hasil transfeksi ditetaskan pada stoples berisi air laut sebanyak 2 L yang diletakkan pada waterbath. Hasil penelitian menunjukkan bahwa gen antivirus telah berhasil diintroduksi ke embrio udang windu. Hasil analisis ragam menunjukkan bahwa perbedaan konsentrasi DNA (5-15 µg tidak berpengaruh nyata (P>0,05 terhadap daya tetas embrio udang windu. Analisis ekspresi gen pada larva udang windu juga menunjukkan adanya aktivitas ekspresi gen antivirus pada semua perlakuan konsentrasi DNA, di mana ekspresi gen antivirus pada larva transgenik lebih tinggi dibandingkan dengan kontrol (tanpa transfeksi. Sintasan pasca-larva PL-1 yang didapatkan pada penelitian ini adalah 12,0%; 10,0%; 10,6%; 12,3%; dan 14,2% masing-masing untuk perlakuan konsentrasi plasmid DNA 5 µg, 10 µg, 15 µg, kontrol positif dan negatif, di mana

  13. KARAKTERISTIK SEKUEN cDNA PENGKODE GEN ANTI VIRUS DARI UDANG WINDU, Penaeus monodon

    Directory of Open Access Journals (Sweden)

    Andi Parenrengi

    2016-11-01

    Full Text Available Transgenesis pada ikan merupakan sebuah teknik modern yang berpotensi besar dalam menghasilkan organisme yang memiliki karakter lebih baik melalui rekombinan DNA gen target termasuk gen anti virus dalam peningkatan resistensi pada udang. Gen anti virus PmAV (Penaeus monodon Anti Viral gene merupakan salah satu gen pengkode anti virus yang berasal dari spesies krustase. Penelitian ini dilakukan untuk mengetahui karakteristik gen anti virus yang diisolasi dari udang windu, Penaeus monodon. Isolasi gen anti virus menggunakan metode Polymerase Chain Reaction (PCR dan selanjutnya dipurifikasi untuk sekuensing. Data yang dihasilkan dianalisis dengan program Genetyx Versi 7 dan basic local alignment search tool (BLAST. Hasil penelitian menunjukkan bahwa gen anti virus PmAV yang berhasil diisolasi dari cDNA udang windu dengan panjang sekuen 520 bp yang mengkodekan 170 asam amino. BLAST-N menunjukkan tingkat similaritas yang sangat tinggi (100% dengan gen anti virus yang ada di GeneBank. Komposisi asam amino penyusun gen anti virus yang paling besar adalah serin (10,00%, sedangkan yang terkecil adalah asam amino prolin dan lisin masing-masing 1,76%. Analisis sekuen gen dan deduksi asam amino (BLAST-P memperlihatkan adanya C-type lectin-like domain (CTLD yang memiliki kemiripan dengan gen C-type lectin yang diisolasi dari beberapa spesies krustase. Transgenic fish technology is a potential modern technique in producing better character organism through DNA recombinant of target genes including anti viral gene for improvement of shrimp immunity. PmAV (Penaeus monodon Anti Viral gene is one of anti viral genes isolated from crustacean species. The research was conducted to analyze the characteristics anti viral gene isolated from tiger prawn, Penaeus monodon. Anti viral gene was isolated using Polymerase Chain Reaction (PCR technique and then purified for sequencing. Data obtained were analyzed using Genetyx Version 7 software and basic local alignment

  14. International Expert - OECD/NEA

    International Nuclear Information System (INIS)

    Yi, Yong Sun

    2009-11-01

    This report was prepared to describe the activities of Yongsun Yi as a technical secretary for the Generation IV VHTR (Very High Temperature Reactor) system. The contents of the report are; i) the GIF (Generation IV International Forum); ii) the GIF governance and technical secretariat; iii) the brief description of the VHTR system iv) the activities of Yongsun Yi as the technical secretary for the VHTR system

  15. Oxochloroalkoxide of the Cerium (IV and Titanium (IV as oxides precursor

    Directory of Open Access Journals (Sweden)

    Machado Luiz Carlos

    2002-01-01

    Full Text Available The Cerium (IV and Titanium (IV oxides mixture (CeO2-3TiO2 was prepared by thermal treatment of the oxochloroisopropoxide of Cerium (IV and Titanium (IV. The chemical route utilizing the Cerium (III chloride alcoholic complex and Titanium (IV isopropoxide is presented. The compound Ce5Ti15Cl16O30 (iOPr4(OH-Et15 was characterized by elemental analysis, FTIR and TG/DTG. The X-ray diffraction patterns of the oxides resulting from the thermal decomposition of the precursor at 1000 degreesC for 36 h indicated the formation of cubic cerianite (a = 5.417Å and tetragonal rutile (a = 4.592Å and (c = 2.962 Å, with apparent crystallite sizes around 38 and 55nm, respectively.

  16. Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

    Directory of Open Access Journals (Sweden)

    Chihyung Kim

    2016-04-01

    Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  17. Deferribacter thermophilus gen. nov., sp. nov., a novel thermophilic manganese- and iron-reducing bacterium isolated from a petroleum reservoir.

    Science.gov (United States)

    Greene, A C; Patel, B K; Sheehy, A J

    1997-04-01

    A thermophilic anaerobic bacterium, designated strain BMAT (T = type strain), was isolated from the production water of Beatrice oil field in the North Sea (United Kingdom). The cells were straight to bent rods (1 to 5 by 0.3 to 0.5 microns) which stained gram negative. Strain BMAT obtained energy from the reduction of manganese (IV), iron(III), and nitrate in the presence of yeast extract, peptone, Casamino Acids, tryptone, hydrogen, malate, acetate, citrate, pyruvate, lactate, succinate, and valerate. The isolate grew optimally at 60 degrees C (temperature range for growth, 50 to 65 degrees C) and in the presence of 2% (wt/vol) NaCl (NaCl range for growth, 0 to 5% [wt/vol]). The DNA base composition was 34 mol% G + C. Phylogenetic analyses of the 16S rRNA gene indicated that strain BMAT is a member of the domain Bacteria. The closest known bacterium is the moderate thermophile Flexistipes sinusarabici (similarity value, 88%). Strain BMAT possesses phenotypic and phylogenetic traits that do not allow its classification as a member of any previously described genus; therefore, we propose that this isolate should be described as a member of a novel species of a new genus, Deferribacter thermophilus gen. nov., sp. nov.

  18. Spectroscopy and chemistry of uranium IV

    International Nuclear Information System (INIS)

    Folcher, G.; Rigny, P.

    1980-06-01

    Different fundamental research papers on uranium IV are presented, some were never edited. Molecular spectroscopy was used for identification and structural study of uranium IV in aqueous or organic solutions. The fields studied are: coordination, stereochemistry, electronic structure and chemical properties. For interpretation of results some studies were made with solid compounds or with thorium compounds or thorium complexes. Knowledge of actinides chemistry is improved, uranium and thorium being models for 5 f ions, extractive chemistry is better understood and new applications are possible [fr

  19. Vectorization at the KENO-IV code

    International Nuclear Information System (INIS)

    Asai, K.; Higuchi, K.; Katakura, J.

    1986-01-01

    The multigroup criticality safety code KENO-IV has been vectorized and tested on the FACOM VP-100 vector processor. At first, the vectorized KENO-IV on a scalar processor was slower than the original one by a factor of 1.4 because of the overhead introduced by vectorization. Making modifications of algorithms and techniques for vectorization, the vectorized version has become faster than the original one by a factor of 1.4 on the vector processor. For further speedup of the code, some improvements on compiler and hardware, especially on addition of Monte Carlo pipelines to the vector processor, are discussed

  20. Functions in Free-Format RPG IV

    CERN Document Server

    Martin, Jim

    2009-01-01

    Written especially for programmers adopting a free-format style, this manual explores the role of functions in writing RPG IV programs. Demonstrating the potential of functions, many topics are explored such as details about existing RPG IV built-in functions, writing new functions, using ILE concepts to use C functions, and utilizing IBM API's functions. Explaining how to write small programs, either as sub-procedures or modules, and how to gather those parts together to make programs that are easy to write and maintain, this is a natural next step for programmers familiar with a free-format