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Sample records for dragon htr graphite

  1. Recent work on graphite corrosion in dragon HTR

    International Nuclear Information System (INIS)

    Wilkinson, V.J.; Parsons, P.D.; Lind, R.

    1976-01-01

    Recent studies are described of graphite corrosion in the Dragon reactor as a consequence of a programme of moisture additions to the helium coolant. The pattern of oxidation was significantly different from that expected from out-of-pile studies. Explanations are suggested in terms of flow and pore structure effects. (orig.) [de

  2. The Dragon project and high temperature reactor (HTR position)

    International Nuclear Information System (INIS)

    Shepherd, L.

    1981-01-01

    After introduction describing the initiation of HTR work at AERE and in West Germany and the USA, the subject is discussed in detail under the headings: the Dragon Reactor Experiment (design and objectives); fuel elements and graphite (description of cooperative research programmes; development of coated fuel particles); helium technology; other Dragon activities. (U.K.)

  3. STIGMA STIG STEGT STAGT STABA, Stress Analysis of Dragon HTR Graphite Structure

    International Nuclear Information System (INIS)

    Kinkead, A.N.

    2002-01-01

    1 - Nature of the physical problem solved: Stress analysis of graphite structures for the DRAGON high temperature reactor is performed by this family of computer codes. Two-dimensional plane strain irradiation dose dependent core problems have been solved. 2 - Method of solution: STAGT, which is the oldest in this series of programmes, can handle multiply connected regions but is confined to plane strain in x-y geometry. Variations in temperature loading during irradiation is accounted for (Wigner strain component.) STIG, is a version of STAGT where an anisotropic elasticity matrix has been introduced to handle transversely isotropic materials. An additional feature of 'STIG' is the introduction of a boundary restraint condition of practical importance to prismatic gas cooled reactor core construction. This is defined as rotational plane strain in which free distortion of the prism arising from overall gradient of temperature and/or fast neutron damage flux coincident with any single direction may be assumed to occur if variation of thermal expansion coefficient with irradiation is included. 'STIGMA' is intended for evaluation of stress and displacement in composite axisymmetrical bodies subject to variable loadings in the axial and radial directions. The code has been prepared to take account of transverse isotropy in material characteristics for up to four separate bonded interface zones within a single composite material problem. Although specifically designed for the analysis of graphite structural components in the fast neutron irradiation environment of a reactor core, it is equally applicable to initial state design of prestressed concrete pressure vessels and other problems involving rotational symmetry. 'STABA'-stress,temperature and bowing analysis. The aim of this quasi 3-D computer code is to apply the principle of rotational plane strain over the full length of a prismatic core component, taking into account spatial variations in fast neutron and

  4. Studies on the behavior of graphite structures irradiated in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R.; Graham, L. W.; Ridealgh, F.

    1971-11-15

    Design data for the physical and mechanical property changes which occur in graphite structural and fuel body components irradiated in an HTR are largely obtained from small specimens tested in the laboratory and in materials test reactors. A brief data summary is given. This graphite physics data can be used to predict dimensional changes, internal stress generation and strength changes in the graphite materials of HTR fuel elements irradiated in the Dragon Reactor. In this paper, the results which have been obtained from post-irradiation examination of a number of fuel pins, are compared with prediction.

  5. An evaluation of the results of the HTR fuel programme conducted in the Dragon reactor experiment

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1982-01-01

    The Dragon Reactor Experiment was used over a period of ten years to investigate the behaviour of HTR fuel elements under realistic service conditions. The purpose of the work was to develop fuel capable of meeting the requirements of commercial power reactors. The studies divided into areas concerned with the mechanical behaviour of the graphite core structure under fast neutron irradiation and the ability of the coated particle fuel to retain fissile products over commercially viable life-cycles. (author)

  6. Approach to equilibrium calculations for the dragon HTR design

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1971-06-10

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

  7. Fuel cycle studies for the Dragon HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A; Nunn, R M; Twitchin, A E

    1971-02-15

    This note reports the progress made at B.N.L. in the study of the fuel cycle for the HTR design described by Daub (1970). The primary purpose of the study is to examine the special problems of the approach to equilibrium fuel cycle.

  8. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  9. Studies on equilibrium fuel management schemes on the Dragon HTR core design

    Energy Technology Data Exchange (ETDEWEB)

    Daub, J; Pedersen, J

    1971-02-03

    The Dragon Project has recently started investigations on fuel management in HTR's with the assumed Dragon design. The study covers the results of investigations into a number of equilibrium fuel management schemes with the 1-dimensional FLATTER code and calculations of the corresponding total power generating costs with the programme TECO.

  10. Design Procedure of Graphite Components by ASME HTR Codes

    International Nuclear Information System (INIS)

    Kang, Ji-Ho; Jo, Chang Keun

    2016-01-01

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet

  11. Design Procedure of Graphite Components by ASME HTR Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ji-Ho; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet.

  12. Nuclear graphite wear properties and estimation of graphite dust production in HTR-10

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Xiaowei, E-mail: xwluo@tsinghua.edu.cn; Wang, Xiaoxin; Shi, Li; Yu, Xiaoyu; Yu, Suyuan

    2017-04-15

    Highlights: • Graphite dust. • The wear properties of graphite. • Pebble bed. • High Temperature Gas-cooled Reactor. • Fuel element. - Abstract: The issue of the graphite dust has been a research focus for the safety of High Temperature Gas-cooled Reactors (HTGRs), especially for the pebble bed reactors. Most of the graphite dust is produced from the wear of fuel elements during cycling of fuel elements. However, due to the complexity of the motion of the fuel elements in the pebble bed, there is no systematic method developed to predict the amount the graphite dust in a pebble bed reactor. In this paper, the study of the flow of the fuel elements in the pebble bed was carried out. Both theoretical calculation and numerical analysis by Discrete Element Method (DEM) software PFC3D were conducted to obtain the normal forces and sliding distances of the fuel elements in pebble bed. The wearing theory was then integrated with PFC3D to estimate the amount of the graphite dust in a pebble bed reactor, 10 MW High Temperature gas-cooled test Reactor (HTR-10).

  13. High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Manzel, R.; Everett, M. R.; Graham, L. W.

    1971-05-15

    The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 1021 n cm-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.

  14. Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

    2015-11-15

    Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing

  15. Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Anthony A. [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)

    2013-07-01

    The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)

  16. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  17. Study on "1"4C content in post-irradiation graphite spheres of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Pi Yue; Xie Feng; Li Hong; Cao Jianzhu

    2014-01-01

    Since the production mechanism of the "1"4C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of "1"4C in fuel elements and graphite spheres of HTR-10, the production mechanism of the "1"4C in graphite spheres was studied. The production sources of the "1"4C in graphite spheres and fuel elements were summarized, the amount of "1"4C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the "1"4C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

  18. On the enhancement of corrosion in HTR graphite due to machining

    International Nuclear Information System (INIS)

    Hatton, D.; Harris, A.M.; Hall, J.A.

    1975-09-01

    A correlation is reported between bands of corrosion on the coolant surface of an inner sleeve from a Dragon reactor tubular interacting pin, and peaks in the axial profile of 60 Co along the pin. Tungsten has been observed at the corroded positions and it is postulated that these materials, deposited from a tungsten carbide tipped tool during machining, cause catalytic corrosion of the graphite. (author)

  19. Achievements of the Dragon Project

    International Nuclear Information System (INIS)

    Rennie, C.A.

    1978-01-01

    The Dragon High Temperature Reactor (HTR) Project began 1 April 1959 under OECD auspices. Extensions in time and budget allowed the project to continue 17 years at a total cost of nearly 100 million dollars under efficient and flexible international management. The reactor design evolved from purged elements and continuously decontaminated helium coolant in a highly contaminated circuit with double containment, to coated particle fuel elements that kept the coolant activity low and permitted easy maintenance. Some difficulties arose from corrosion of heat exchangers and stainless steel pipes and from dimensional changes in the reflector graphite. These problems were easily solved. Some ten years of experimental operation were very successful and demonstrated the soundness of the concept. The Dragon reactor proved to be a very useful test bed for a number of different HTR of different HTR fuel element concepts. (author)

  20. Irradiation creep in reactor graphites for HTR applications. [Neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Veringa, H J; Blackstone, R [Stichting Reactor Centrum Nederland, Petten

    1976-01-01

    A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400/sup 0/C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400/sup 0/C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density.

  1. The irradiation creep in reactor graphites for HTR applications

    International Nuclear Information System (INIS)

    Veringa, H.J.; Blackstone, R.

    1976-01-01

    A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400 0 C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400 0 C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density. (author)

  2. Status of IAEA international data base on irradiated graphite properties with respect to HTR engineering issues

    International Nuclear Information System (INIS)

    Hacker, P.J.; Haag, G.

    2002-01-01

    The International Database on Irradiated Nuclear Graphite Properties contains data on the physical, chemical, mechanical and other relevant properties of graphites. Its purpose is to provide a platform that makes these properties accessible to approved users in the fields of nuclear power, nuclear safety and other nuclear science and technology applications. The database is constructed using Microsoft Access 97 software and has a controlled distribution by CD ROM to approved users. This paper describes the organisation and management of the database through administrative arrangements approved by the IAEA. It also outlines the operation of the database. The paper concludes with some remarks upon and illustrations of the usefulness of the database for the design and operation of HTR. (authors)

  3. Application of the Dragon reactor experiment to the safety evaluation of current HTR systems

    International Nuclear Information System (INIS)

    Ashworth, F.P.O.; Faircloth, R.L.

    1976-01-01

    An important component of the confidence required for the safety assessment of high-temperature reactors is the experimental proof of phenomena such as fission product release or core corrosion. The most convincing experiments are those carried out in a reactor. This paper outlines the scope of experiments relevant to safety which can be done in the Dragon Reactor Experiment and describes as an example the experimental campaign and the current outcome of the work on validating the predictions of caesium release and migration. (author)

  4. A 3-D inelastic analysis of HTR graphite structures and a comparison with A 2-D approach

    International Nuclear Information System (INIS)

    Willaschek, J.

    1979-01-01

    In High Temperature Reactor Cores (HTR) a large number of elements are constructed of nuclear graphite. The dimensions of the graphite components are limited by stresses and strains resulting from thermal loads, irradiation induced dimensional changes and stress-dependent irradiation creep. Therefore it is necessary to examine the feasibility of design concepts with regard to the structural integrity of the material. This paper presents an analysis of a radial reflector concept for use in a 3000 MWth HTR for process heat production. This concept of a pebble bed reactor (OTTO cycle) requires reflector dimensions and shapes which have previously not been used and which may exceed acceptable stress limits. Graphite reflector elements in a HTR are subject to a high fluence of fast neutrons. The fluence varies spatially within an element. Irradiation-induced strains occur which in turn vary non-linearly with the fluence. At low fluences the graphite shrinks. With increasing fluence shrinkage is saturated and after a 'turn-around' point the graphite begins to swell. The net effect of fluence gradient and irradiation-induced strain is a 'necking' of the element which moves radially outwards with time. In this paper a three-dimensional inelastic analysis of a graphite block with the above deformation history is described. The influence of irradiation on dimensional stability and other material properties was taken into account. Numerical results were obtained with the finite-element computer code ADINA, modified at INTERATOM for the task in hand. The radial reflector block was modelled using 21-node three-dimensional continuum elements of elastic-creep material. The element stiffness matrices were calculated using the standard 2x2x2 Gauss integration; material nonlinearities with quadratic displacement functions and linearised initial strains were employed. (orig.)

  5. Dragon.

    Directory of Open Access Journals (Sweden)

    Horacio Ortiz

    2006-06-01

    Full Text Available L’ouvrage d’Howard Becker, The Tricks of the Trade , finit sur une parabole, que l’auteur nous indique comme étant une métaphore sur la notion d’illumination. Cette métaphore est ce qui se rapproche le plus, selon Becker, du fait d’avoir, jusqu’à l’os, une manière de penser qui est celle des sciences sociales. La métaphore sur l’illumination particulière des sciences sociales compare les chercheuses en sciences sociales à des dragons océaniques. Pour celles 1 qui ...

  6. The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

    International Nuclear Information System (INIS)

    Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

    2014-01-01

    Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

  7. Temperature coefficients in the Dragon low-enriched power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1972-05-15

    The temperature coefficient of the fuel and of the moderator have been evaluated for the Dragon HTR design for different stages in reactor life, initial core, end of no-refuelling period and equilibrium conditions. The investigation has shown the low-enriched HTR to have a strong, positive moderator coefficient. In some cases and for special operating conditions, even leading to a positive total temperature coefficient. This does not imply, however, that the HTR is an unsafe reactor system. By adequate design of the control system, safe and reliable operating characteristics can be achieved. This has already been proved satisfactory through many years of operation of other graphite moderated systems, such as the Magnox stations.

  8. Criteria for the selection of graphites for HTR integral block fuel elements

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1980-01-01

    This paper is concerned with the special requirements for integral block fuel elements of the type first used in the Fort St. Vrain reactor. The main idea of these elements is that the carrier block and separate graphite clad fuel pins are combined into a single monolith. This combination leads to lower fabrication costs and some improvement in the thermal performance (lower temperature difference between fuel and the surface of heat transfer into the coolant). The advent of block fuel for HTRs of the Fort St. Vrain type has placed a fresh emphasis on the selection of graphite for block manufacture in respect of physical properties. This is because the temperature distributions typical of such fuelled blocks lead to shutdown stresses close to the maximum the graphite can sustain without damage. Figures presented in this paper suggest that the physical properties of the graphite can play a relatively large part in reducing such stress levels and that guidance on the key requirements for suitable specifications is therefore particularly needed by the manufacturers of fuel block graphites. While graphites for fuel blocks have this special need for combinations of physical properties which lead to low thermal and shrinkage stresses, the other characteristics must also receive attention. A low graphite cost combined with good homogeneity in the brick, so that waste minimized, are still necessary, while isotropy is also very important

  9. Study on thermal conductivity of HTR spherical fuel element matrix graphite

    International Nuclear Information System (INIS)

    Zhang Kaihong; Liu Xiaoxue; Zhao Hongsheng; Li Ziqiang; Tang Chunhe

    2014-01-01

    Taking the spherical fuel element matrix graphite ball samples as an example, this paper introduced the principle and method of laser thermal conductivity meter, as well as the specific heat capacity, and analyzed the effects of different test methods and sampling methods on the thermal conductivities at 1000 ℃ of graphite material. The experimental results show that the thermal conductivities of graphite materials tested by synchronous thermal analyzer combining with laser thermal conductivity meter were different from that directly by laser thermal conductivity meter, the former was more reliable and accurate than the later; When sampling from different positions, central samples had higher thermal conductivities than edging samples, which was related to the material density and porosity at the different locations; the thermal conductivities had obvious distinction between samples from different directions, which was because the layer structure of polycrystalline graphite preferred orientation under pressure, generally speaking, the thermal conductivities perpendicular to the molding direction were higher than that parallel to the molding direction. Besides this, the test results show that the thermal conductivities of all the graphite material samples were greater than 30 W/(m (K), achieving the thermal performance index of high temperature gas cooled reactor. (authors)

  10. HTR-TN achievements and prospects for future developments

    International Nuclear Information System (INIS)

    Hittner, D.; Angulo, C.; Basini, V.; Bogusch, E.; Breuil, E.; Buckthorpe, D.; Chauvet, V.; Futterer, M.A.; Van Heek, A.; Von Lensa, W.; Yvon, P.

    2011-01-01

    It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the 'Strategic Energy Technology Plan (SET-Plan)' recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to

  11. IRPhE-HTR-ARCH-01, Archive of HTR Primary Documents

    International Nuclear Information System (INIS)

    2004-01-01

    Description: High Temperature Reactor Studies, including experiments in critical facilities or in prototypes have been carried out in the past. Information gathered, experience gained and experimental data produced are of value for the development of future advanced HTRs. For the purpose of knowledge, competence, information preservation and management, computer readable archives have been established. The present archive includes several relevant documents relative to the following: - Graphite Moderated Critical Facility, CESAR at Cadarache. Dragon Countries Physics Meetings (DCPM); - OTTO Pebble Bed Reactors; - Gulf - HTGR Experiments; - Zero Power MARIUS Reactor; - Pebble-bed KAHTER Critical Facility; - Helium Cooled Fast Reactor Assessment Studies; - Gas Cooled Reactor Technology Safety and Siting; - Initial Evaluation of the Gas-Turbine Modules HTGCR; - A report on Nuclear Graphite; - AVR Reactor Juelich (new in version 02); - HTR IAEA proceedings (new in version 02); - Studies at IRI Delft(new in version 02); - Studies and experiments at PSI Villigen (new in version 02); 2 - Related or auxiliary information: IRPHE-DRAGON-DPR, high Temperature Reactor Dragon Project, Primary Documents NEA-1726/01. 3 - Software requirements: Acrobat Reader, Microsoft Word, HTML Browser required

  12. Strategy Study on Treatment and Disposal of the Radioactive Graphite Waste of HTR

    International Nuclear Information System (INIS)

    Li Junfeng; Ma Tao; Wang Jianlong

    2014-01-01

    The possible options to change HTGR spent fuel into an acceptable form for repository disposal were discussed. The progresses of physical, chemical, and electrochemical separation of graphite from the HTGR fuel elements were reviewed. The advantages and disadvantages of each method were listed out. The total waste volume of each method was compared. The preferred option depends on the waste acceptance criteria for the repository, availability of low level waste disposal for graphite, overall economics, and overall risks. The minimum processing that yields an acceptable waste form also gives the lowest costs as well as the simplest process and the least risk. The options that could be used for treating HTGR spent fuels were listed out. The strategy for treating HTGR spent fuels and the packages needed for repository were discussed. (author)

  13. Development of a dry-mechanical graphite separation process and elimination of the separated carbon for the reprocessing of spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Kronschnabel, H.

    1982-01-01

    Due to the C-14 distribution the separation of the particle-free outer region of the spherical HTR fuel element with subsequent solidification of the separated carbon makes it possible to reduce by half the remaining C-14 inventory in the inner particle region to be further treated. Separation of the particle-free outer region by a newly developed sphere-peeling milling machine, conditioning the graphite into compacts and in-situ cementation into a salt-mine are the basic elements of this head-end process variation. An annual cavern volume of approx. 2000 m 3 will be needed to ultimately store the graphite of the particle-free outer region, which corresponds to a reprocessing capacity of 50 GWsub(e) installed HTR power. The brush-disintegration of the remaining inner particle region and the resulting peel-brush-preparation are capable of separating 95% of the graphite without any heavy metal losses. With the mentioned reprocessing capacity an annual cavern volume of approx. 16.500 m 3 is required. (orig.) [de

  14. Computation of deformations and stresses in graphite blocks for HTR core survey purposes

    International Nuclear Information System (INIS)

    Besdo, Dieter; Theymann, W.

    1975-01-01

    Stresses and deformations in graphite fuel elements for HTRs are caused by the temperature distribution and by irradiation under influence of creep, shrinking, thermal strains, and elastic deformations. The global deformations and the stress distribution in a prismatic fuel-element containing regularly distributed axial holes for the coolant flow and the fuel sticks, can be computed in the following manner: the block with its holes is treated as an effective homogeneous continuum with an equivalent global behaviour. Assuming that the fourth-order-tensor of the elastic constants is proportional to the corresponding tensor in the constitutive equations for creep, only the effective strains are of interest. The values of temperature and dose may be given in n points of the block at certain points of time. Then, the inelastic nonthermal strains are integrated by a Runge-Kutta-procedure in the n points. When interpolated and combined with thermal strains, they are incompatible. Hence, they produce elastic deformations which cause creep and can be computed by use of a Ritz-polynomial-series by help of a specific principle of the minimum of potential energy. Excessive computing time can be avoided easily since the influence of the local variation of the elastic constants within the block is almost negligible and, therefore, of practically no importance for the determination of the elastic strains. By this reason some matrices can be calculated a priori, and the elastic deformations are obtained by multiplications of these matrices rather than inversions. Therefore, this method is particularly suited for the computation of deformations and stresses for reactor core survey purposes where a large number (up to 7000 blocks) have to be treated

  15. OECD high temperature reactor project Dragon

    International Nuclear Information System (INIS)

    1975-01-01

    Information is presented concerning the Dragon reactor support studies and fuel irradiation programs, HTGR and fuel graphite studies, primary circuit materials, reactor safety evaluation, and administration

  16. French programme for HTR fuel

    International Nuclear Information System (INIS)

    Gillet, R.M.

    1991-01-01

    It is reported that in the frameworks of the French HTR research program, stopped in 1979 the HTR coated particle fuel, fuel rod and prismatic fuel element design have been successfully developed and irradiation tested in France and specific examination methods for irradiated fuel particles, rods and graphite blocks have been developed. Currently CEA is involved in fission product transport experiments sponsored by the US Department of Energy and performed in the COMEDIE loop. Finally the CEA follows progress and developments in HTR fuel research and development throughout the world. 1 tab

  17. Physics experiment on the Dragon reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, C.

    1974-10-15

    The paper describes a set of DRAGON experiments planned to measure burn-up effects in DRAGON irradiated fuel. Irradiated fuel elements from DRAGON are to be subjected to reactivity measurements in the HECTOR experimental reactor to infer the residual U235 content followed by isotopic analyses at CEA laboratories in 1975. Fast neutron damage to DRAGON graphite is compared to fast neutron dose measurements using Ni58 (n,p) Co58 activation wires in both DRAGON and the DIDO MTR. Gamma scanning of irradiated fuel elements are used to compare axial power profiles to those derived from two-dimensional and three-dimensional calculations of the DRAGON reactor.

  18. Effects of reaction temperature and inlet oxidizing gas flow rate on IG-110 graphite oxidation used in HTR-PM

    International Nuclear Information System (INIS)

    Sun Ximing; Dong Yujie; Zhou Yangping; Shi Lei; Sun Yuliang; Zhang Zuoyi; Li Zhengcao

    2017-01-01

    The oxidation behavior of a selected nuclear graphite (IG-110) used in Pebble-bed Module High Temperature gas-cooled Reactor was investigated under the condition of air ingress accident. The oblate rectangular specimen was oxidized by oxidant gas with oxygen mole fraction of 20% and flow rates of 125–500 ml/min at temperature of 400–1200°C. Experiment results indicate that the oxidation behavior can also be classified into three regimes according to temperature. The regime I at 400–550°C has lower apparent activation energies of 75.57–138.59 kJ/mol when the gas flow rate is 125–500 ml/min. In the regime II at 600–900°C, the oxidation rate restricted by the oxygen supply to graphite is almost stable with the increase of temperature. In the regime III above 900°C, the oxidation rate increases obviously with the increase of temperature.With the increase of inlet gas flow from 125 to 500 ml/min, the apparent activation energy in regime I is increased and the stableness of oxidation rate in regime II is reduced. (author)

  19. Present status of research and development for HTR in China

    Energy Technology Data Exchange (ETDEWEB)

    Dazhong, Wang; Daxin, Zhong; Yuanhul, Xu [Institute of Nuclear Energy Technology, Tsinghua University, Beijing (China)

    1990-07-01

    The HTR R and D Project is being carried out in the relevant institutions in China. Some topics are covered such as, fuel element technology, graphite development, fuel element handling system, helium technology, fuel reprocessing technology as well as HTR design study. Some results of HTR research work are described. In addition, to provide a test facility for investigation of HTR Module reactor safety and process heat application of HTR, a joint project on building a 10 MW test HTR with Siemens-Interatom, KFA Juelich and INET is going on. The conceptual design of 10 MW test HTR has been completed by the joint group. In parallel the application study of HTR Module is being carried out for the oil industry, petrochemical industry as well as power generation. Some preliminary results of the application study, for example, for heavy oil recovery on Shengli oil field and process heat application in Yan shan petroleum company, are described. (author)

  20. Graphite

    Science.gov (United States)

    Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

  1. Gas cooled HTR

    International Nuclear Information System (INIS)

    Schweiger, F.

    1985-01-01

    In the He-cooled, graphite-moderated HTR with spherical fuel elements, the steam generator is fixed outside the pressure vessel. The heat exchangers are above the reactor level. The hot gases stream from the reactor bottom over the heat exchanger, through an annular space around the heat exchanger and through feed lines in the side reflector of the reactor back to its top part. This way, in case of shutdown there is a supplementary natural draught that helps the inner natural circulation (chimney draught effect). (orig./PW)

  2. The Dragon Project origins, achievements and legacies

    International Nuclear Information System (INIS)

    Price, M.S.T.

    2012-01-01

    The lineage of the Dragon Project can be traced back to 1955 when the United Kingdom launched a nuclear power programme which involved the construction of large graphite moderated reactors fuelled with natural uranium and cooled by carbon dioxide. Not long afterwards the European Nuclear Energy Agency (ENEA) of the then newly formed Organisation for European Economic Cooperation (OEEC), in the spirit of the time, sought to encourage the construction of nuclear power stations and the development of joint nuclear undertakings. The United Kingdom Atomic Energy Research Establishment (AERE) had, since 1949, been studying possible long term improvements in energy conversion efficiency resulting from higher coolant gas temperatures and the use of ceramic materials. A 1955 paper on gas-cooled reactors using the U-233/thorium cycle attracted interest and this progressed to the definition of an initial programme. The high temperature work led to a proposal for a 20 MW(Th) Reactor Experiment and one important consequence of the ENEA/OEEC initiative was the setting up in April 1959 of the international Dragon Project Agreement. Initial experiments at Harwell in 1957 had involved the coating of small spheroidal particles of uranium carbide or oxide with pyrolytic carbon which were then bonded with carbonaceous material. But experiments demonstrated that fission products such as caesium, strontium or barium could diffuse through such coatings. This led in 1961 to the modification of the coated particle design by the addition of an intermediate layer of silicon or zirconium carbide. The small size of the particles necessitated a statistical approach to quality during manufacture and effort was concentrated on the minimisation of the broken or defective particle fraction. The subsequent operation of the Dragon Reactor for over 10 years confirmed the benign nature of a HTR. It also proved that fuel bodies made with coated particles were capable of maintaining a high degree of

  3. Prospective studies of HTR fuel cycles involving plutonium

    International Nuclear Information System (INIS)

    Bonin, B.; Greneche, D.; Carre, F.; Damian, F.; Doriath, J.Y.

    2002-01-01

    High Temperature Gas Cooled reactors (HTRs) are able to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This flexibility is due to an uncoupling between the parameters of cooling geometry, and the parameters which characterize neutronic optimisation (moderation ratio or heavy nuclide concentration and distribution). Among other advantageous features, an HTR core has a better neutron economy than a LWR because there is much less parasitic capture in the moderator (capture cross section of graphite is 100 times less than the one of water) and in internal structures. Moreover, thanks to the high resistance of the coated particles, HTR fuels are able to reach very high burn-ups, far beyond the possibilities offered by other fuels (except the special case of molten salt reactors). These features make HTRs especially interesting for closing the nuclear fuel cycle and stabilizing the plutonium inventory. A large number of fuel cycle studies are already available today, on 3 main categories of fuel cycles involving HTRs : i) High enriched uranium cycle, based on thorium utilization as a fertile material and HEU as a fissile material; ii) Low enriched uranium cycle, where only LEU is used (from 5% to 12%); iii) Plutonium cycle based on the utilization of plutonium only as a fissile material, with (or without) fertile materials. Plutonium consumption at high burnups in HTRs has already been tested with encouraging results under the DRAGON project and at Peach Bottom. To maximize plutonium consumption, recent core studies have also been performed on plutonium HTR cores, with special emphasis on weapon-grade plutonium consumption. In the following, we complete the picture by a core study for a HTR burning reactor-grade plutonium. Limits in burnup due to core neutronics are investigated for this type of fuel. With these limits in mind, we study in some detail the Pu cycle in the special case of a

  4. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    International Nuclear Information System (INIS)

    Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

    1985-01-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e. fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations

  5. Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

    Energy Technology Data Exchange (ETDEWEB)

    Reitsamer, G; Proksch, E; Stolba, G; Strigl, A; Falta, G; Zeger, J [Department of Chemistry, Austrian Research Centre Seibersdorf Ltd., Seibersdorf (Austria)

    1985-07-01

    Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e., fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations.

  6. The Dragon reactor experiment

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

  7. Sea Dragon

    National Research Council Canada - National Science Library

    1997-01-01

    .... In preparation for these changes, the Navy is exploring new command and control relationships, and the Marine Corps established Sea Dragon to experiment with emerging technologies, operational...

  8. The physics design of the HTR-1160

    International Nuclear Information System (INIS)

    Huebner, A.; Brandes, S.

    1975-01-01

    This paper describes the physica design of the reactor core of the helium cooled, graphite moderated high-temperature reactor HTR-1160. A discussion is given of the design criteria, the calculational methods, the burnup cycle, the power distribution and the reactivity control. (orig.) [de

  9. Wave Dragon

    DEFF Research Database (Denmark)

    Tedd, James; Kofoed, Jens Peter; Friis-Madsen, Erik

    2008-01-01

    Since March 2003 a prototype of Wave Dragon has been tested in an inland sea in Denmark. This has been a great success with all subsystems tested and improved through working in an offshore environment. The project has proved the Wave Dragon device and has enabled the next stage, a production sized...

  10. Wave Dragon

    DEFF Research Database (Denmark)

    Kofoed, Jens Peter; Frigaard, Peter; Sørensen, H. C.

    1998-01-01

    This paper concerns with the development of the wave energy converter (WEC) Wave Dragon. This WEC is based on the overtopping principle. An overview of the performed research done concerning the Wave Dragon over the past years is given, and the results of one of the more comprehensive studies, co...

  11. A brief outline of the proposed data/physics calculation scheme proposed for Dragon

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, C.

    1974-10-15

    The paper describes the data handling system COSMOS, that was in current use at the Dounreay Prototype Fast Reactor (PFR), to provide a template for data handling for the Dragon Project wherein the Physics and Engineering Modules would be centered instead around the WIMS-E suite of codes which are more applicable to the HTR. The tailoring of COSMOS to Dragon needs and the adaption of UKAEA and existing Dragon codes to operate on the data interface would require considerable modification.

  12. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2017-01-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  13. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  14. Scripting DRAGON

    International Nuclear Information System (INIS)

    Hebert, A.

    2002-01-01

    The paper describes an user-oriented framework specifically designed to facilitate the use of legacy Fortran codes similar to Dragon. The basic idea is to use a bytecode interpreted language as a glue to link all the components required by the end user. This scripting approach is illustrated with Dragon, where we have replaced the control language CLE-2000 with a bytecode interpreted language, without having to modify Dragon. We have shown how Python or Java can be used to link the Dragon modules together and to construct an object-oriented user interface. Python or Java can also be used to construct execution procedures, calculation schemes and graphical user interfaces. Java was finally selected as the most interesting choice. This approach can be used with other legacy Fortran codes, as soon as their input/output data structures are Dragon-compatible. The only modification required on Fortran code is the replacement of some common blocks by associative tables, already available with the LCM application programming interface. (author)

  15. Fuel management of HTR-10

    International Nuclear Information System (INIS)

    Wu Zongxin; Jing Xingqing

    2001-01-01

    The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

  16. Wave Dragon

    DEFF Research Database (Denmark)

    Kramer, Morten; Frigaard, Peter

    På foranledning af Löwenmark F.R.I, er der udført numeriske beregninger af Wave Dragons (herefter WD) armes effektivitet for forskellige geometriske udformninger. 5 geometriske modeller, hvor WD's arme er forkortet/forlænget er undersøgt for 3 forskellige drejninger af armene. I alt er 15...

  17. Fabrication technology of spherical fuel element for HTR-10

    International Nuclear Information System (INIS)

    He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

    2002-01-01

    R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10 -5

  18. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Lai Shigang; Sun Libin; Zhang Zhengming

    2013-01-01

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  19. Wave Dragon

    DEFF Research Database (Denmark)

    Tedd, James; Kofoed, Jens Peter; Knapp, W.

    2006-01-01

    Wave Dragon is a floating wave energy converter working by extracting energy principally by means of overtopping of waves into a reservoir. A 1:4.5 scale prototype has been sea tested for 20 months. This paper presents results from testing, experiences gained and developments made during this ext......Wave Dragon is a floating wave energy converter working by extracting energy principally by means of overtopping of waves into a reservoir. A 1:4.5 scale prototype has been sea tested for 20 months. This paper presents results from testing, experiences gained and developments made during...... this extended period. The prototype is highly instrumented. The overtopping characteristic and the power produced are presented here. This has enabled comparison between the prototype and earlier results from both laboratory model and computer simulation. This gives the optimal operating point and the expected...... power of the device. The project development team has gained much soft experience from working in the harsh offshore environment. In particular the effect of marine growth in the draft tubes of the turbines has been investigated. The control of the device has been a focus for development as is operates...

  20. Dragons as Amulets, Dragons as Talismans, Dragons as Counselors.

    Science.gov (United States)

    Stevenson, Robert G.

    1994-01-01

    Notes that, in diverse historical and cultural settings, dragons have served as protective amulets/powerful talismans to protect/enhance powers of those who possess them. Explores use of such personal symbols in dealing with personal adversity and suggests methods in which dragon symbol can be used to promote discussion of feelings, problems, and…

  1. IRPhE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents

    International Nuclear Information System (INIS)

    2004-01-01

    Description: The DRAGON Reactor Experiment (DRE): The first demonstration High temperature gas reactor (HTGR) was built in the 1960's. Thirteen OECD countries began a project in 1959 to build an experimental reactor known as Dragon at Winfrith in the UK. The reactor - which operated successfully between 1966 and 1975 - had a thermal output of 20 MW and achieved a gas outlet temperature of 750 deg. C. The High Temperature Reactor concept, if it justified its expectations, was seen as having its place as an advanced thermal reactor between the current thermal reactor types such as the PWR, BWR, and AGR and the sodium cooled fast breeder reactor. It was expected that the HTR would offer better thermal efficiency, better uranium utilisation, either with low enriched uranium fuel or high enriched uranium thorium fuel, better inherent safety and lower unit power costs. In the event all these potential advantages were demonstrated to be in principle achievable. This view is still shared today. In fact Very High Temperature Reactors is one of the concepts retained for Generation IV. Projects on constructing Modular Pebble Bed Reactors are under way. Here all available Dragon Project Reports (DPR) - approximately 1000 - are collected in electronic form. An index points to the reports (PDF format); each table in the report is accessible in EXCEL format with the aim of facilitating access to the data. These reports describe the design, experiments and modelling carried out over a period of 17 years. 2 - Related or auxiliary information: IRPHE-HTR-ARCH-01, Archive of HTR Primary Documents NEA-1728/01. 3 - Software requirements: Acrobat Reader, Microsoft Word, HTML Browser required

  2. HTR Development in China

    International Nuclear Information System (INIS)

    Wang Dazhong

    2014-01-01

    The roles of HTRs in China: 1. Due to the inherent safety features, high efficiency of electricity generation, site flexibility, the modular HTR can act as a supplement to LWR for small and medium size power generation. 2. Co-generation to supply steam up to 600℃, for petroleum refinery, oil sand and oil shale processing, sea water desalination and district heating, etc. 3. Hydrogen production at 900~1000 ℃ by V/HTR. Conclusions and prospects: • China’s energy system will experience transition and reform in the future; • Nuclear energy will play an irreplaceable role in China’s energy development; • Due to the excellent features of inherent safety, the HTR is a promising technology for electricity generation and process heat utilization; • Further international cooperation and exchanges need to be enhanced

  3. Thermodynamic correlations for the accident analysis of HTR's

    International Nuclear Information System (INIS)

    Rehm, W.; Jahn, W.; Finken, R.

    1976-12-01

    The thermal properties of Helium and for the case of a depressurized primary circuit, various mixtures of primary cooling gas were taken into consideration. The temperature dependence of the correlations for the thermal properties of the graphite components in the core and for the structural materials in the primary circuit are extrapolated about normal operation conditions. Furthermore the correlations for the effective thermal conductivity, the heat transfer and pressure drop are described for pebble bed HTR's. In addition some important heat transfer data of the steam generator are included. With these correlations, for example accident sequences with failure of the afterheat removal systems are discussed for pebble bed HTR's. It is concluded that the transient temperature behaviour demonstrates the inherent safety features of the HTR in extreme accidents. (orig.) [de

  4. Preparation of spherical fuel elements for HTR-PM in INET

    International Nuclear Information System (INIS)

    Xiangwen, Zhou; Zhenming, Lu; Jie, Zhang; Bing, Liu; Yanwen, Zou; Chunhe, Tang; Yaping, Tang

    2013-01-01

    Highlights: • Modifications and optimizations in the manufacture of spherical fuel elements (SFE) for HTR-PM are presented. • A newly developed overcoater exhibits good stability and high efficiency in the preparation of overcoated particles. • The optimized carbonization process reduces the process time from 70 h in the period of HTR-10 to 20 h. • Properties of the prepared SFE and matrix graphite balls meet the design specifications for HTR-PM. • In particular the mean free uranium fraction of 5 consecutive batches is only 8.7 × 10 −6 . -- Abstract: The spherical fuel elements were successfully manufactured in the period of HTR-10. In order to satisfy the mass production of fuel elements for HTR-PM, several measures have been taken in modifying and optimizing the manufacture process of fuel elements. The newly developed overcoater system and its corresponding parameters exhibited good stability and high efficiency in the preparation of overcoated particles. The optimized carbonization process could reduce the carbonization time from more than 70 h to 20 h and improve the manufacturing efficiency. Properties of the manufactured spherical fuel elements and matrix graphite balls met the design specifications for HTR-PM. The mean free uranium fraction of 5 consecutive batches was 8.7 × 10 −6 . The optimized fuel elements manufacturing process could meet the requirements of design specifications of spherical fuel elements for HTR-PM

  5. Postirradiation examination of HTR fuel

    International Nuclear Information System (INIS)

    Nabielek, H.; Reitsamer, G.; Kania, M.J.

    1986-01-01

    Fuel for the High Temperature Reactor (HTR) consists of 1 mm diameter coated particles uniformly distributed in a graphite matrix within a cold-molded 60 mm diameter spherical fuel element. Fuel performance demonstrations under simulated normal operation conditions are conducted in accelerated neutron environments available in Material Test Reactors and in real-time environments such as the Arbeitsgemeinschaft Versuchsreaktor (AVR) Juelich. Postirradiation examinations are then used to assess fuel element behavior and the detailed performance of the coated particles. The emphasis in postirradiation examination and accident testing is on assessment of the capability for fuel elements and individual coated particles to retain fission products and actinide fuel materials. To accomplish this task, techniques have been developed which measures fission product and fuel material distributions within or exterior to the particle: Hot Gas Chlorination - provides an accurate method to measure total fuel material concentration outside intact particles; Profile Electrolytic Deconsolidation - permits determination of fission product distribution along fuel element diameter and retrieval of fuel particles from positions within element; Gamma Spectrometry - provides nondestructive method to measure defect particle fractions based on retention of volatile metallic fission products; Particle Cracking - permits a measure of the partitioning of fission products between fuel kernel and particle coatings, and the derivation of diffusion parameters in fuel materials; Micro Gas Analysis - provides gaseous fission product and reactive gas inventory within free volume of single particles; and Mass-spectrometric Burnup Determination - utilizes isotope dilution for the measurement of heavy metal isotope abundances

  6. Initial study on burnable poisons in the Dragon HTR design

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U; Pedersen, J

    1971-06-15

    A first study on the effects of burnable poisons in a High Temperature Reactor is given in this paper, and some of the problems concerning the layout and distribution of burnable poison sticks in the core are explained. Time has not allowed us to obtain satisfactory solutions to these problems, but we hope, that this study could form the basis of valuable discussions on ways and means to overcome the difficulties of burnable poison management in HTRs.

  7. Tritium in HTR systems

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1987-07-01

    Starting from the basis of the radiological properties of tritium, the provisions of present-day radiation protection legislation are discussed in the context of the handling of this radionuclide in HTR plants. Tritium transportation is then followed through from the place of its creation up until the sink, i.e. disposal and/or environmental route, and empirical values obtained in experiments and in plant operation translated into guidelines for plant design and planning. The use of the example of modular HTR plants permits indication that environmental contamination via the 'classical' routes of air and water emissions, and contamination of products, and resulting consumer exposure, are extremely low even on the assumption of extreme conditions. This leads finally to a requirement that the expenditure for implementation of measures for further reduction of tritium activity rates be measured against low radiological effect. (orig.) [de

  8. International HTR activities

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    Asea Brown Boveri AG (ABB) and their subsidiary High Temperature Reactor Construction GmbH (HRB) have brought the pebble bed high temperature reactor to the edge of being ready for the market with the construction and operation of the AVR reactor at Juelich and the THTR 300 at Hamm-Uentrop. Siemens/Interatom have developed the HTR modular concept and, together with their partners HRB, KFA, Rheinbraun Bergbauforschung have taken the nuclear process heat project to its present advanced state of development. The further introduction of the HTR to the market is a long-term objective, due to the present market situation. ABB and Siemens AG have therefore agreed to collaborate by forming a joint company. (orig.)

  9. Long-term testing of HTR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Nickel, H.

    1986-12-01

    The extensive results from irradiation experiments carried out on coated particles, on graphitic matrices of different composition and on integral fuel elements have shown that the spherical fuel elements with high-enriched uranium/thorium mixed-oxide particles and optimized graphitic matrix are available for use in the planned HTR facilities. A concentrated qualification programme is on the way in order to bring the fuel elements with particles from low-enriched uranium dioxide (LEU) and TRISO coating to a comparable level of experience and knowledge, i.e. to make them licensable for the planned HTR facilities. (orig.) [de

  10. Hadron Dragons strike again

    CERN Multimedia

    2009-01-01

    The CERN Dragon Boat team – the Hadron Dragons – achieved a fantastic result at the "Paddle for Cancer" Dragon Boat Festival at Lac de Joux on 6 September. CERN Hadron Dragons heading for the start line.Under blue skies and on a clear lake, the Hadron Dragons won 2nd place in a hard-fought final, following top times in the previous heats. In a close and dramatic race – neck-and-neck until the final 50 metres – the local Lac-de-Joux team managed to inch ahead at the last moment. The Hadron Dragons were delighted to take part in this festival. No one would turn down a day out in such a friendly and fun atmosphere, but the Dragons were also giving their support to cancer awareness and fund-raising in association with ESCA (English-Speaking Cancer Association of Geneva). Riding on their great success in recent competitions, the Hadron Dragons plan to enter the last Dragon Boat festival of 2009 in Annecy on 17-18 October. This will coincide with t...

  11. Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

    International Nuclear Information System (INIS)

    Ortensi, J.; Pope, M.A.; Ferrer, R.M.; Cogliati, J.J.; Bess, J.D.; Ougouag, A.M.

    2010-01-01

    The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INL's current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in the NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2-3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral

  12. HTR characteristics affecting reactor physics

    International Nuclear Information System (INIS)

    Ehlers, K.

    1980-01-01

    A physical description of high-temperature has-cooled reactors is given, followed by an overview of HTR characteristics. The emphasis is placed on the HTR fuel cycle alternatives and thermohydraulics of pebble bed core. Some prospects of HTRs in the Federal Republic of Germany are also presented

  13. Plaster-Wrap Dragons

    Science.gov (United States)

    Vance, Shelly

    2012-01-01

    In this article, the author describes how her students constructed a three-dimensional sculpture of a dragon using plaster wrap and other materials. The dragons were formed from modest means--using only a toilet-paper tube, newsprint, tape and wire.

  14. Chinese New Year Dragons.

    Science.gov (United States)

    Balgemann, Linda

    2000-01-01

    Presents an art project, used in a culturally diverse curriculum, in which second grade students create Chinese New Year dragons. Describes the process of creating the dragons, from the two-week construction of the head to the accordion-folded bodies. (CMK)

  15. Study on the production mechanism of Co-60 in the primary loop of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Xie Feng; Li Hong; Cao Jianzhu; Li Fu; Wei Liqiang

    2015-01-01

    Co-60 is an activated metallic erosion product, which is very important for waste management and decommissioning work of pressurized water reactor (PWR) power plants. Recent measurement on the samples from the primary loop of HTR-10 indicates the existence of Co-60. In current paper, the preliminary experimental results in HTR-10 will be introduced, and the production mechanism of Co-60 in the pebble bed high temperature gas-cooled reactors will be summarized and compared with that in PWRs and Germany High Temperature Nuclear Reactor (AVR). The further experiments with decomposing the post-irradiation graphite spheres of HTR-10 are put forward, which will promote the further study to testify the production sources of Co-60 and be of great significance in the waste minimization and the decommissioning work of HTR-10. (author)

  16. Operating experience with the DRAGON High Temperature Reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.A.; Capp, P.D.

    2002-01-01

    The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

  17. HTR Plans in Poland

    International Nuclear Information System (INIS)

    Sobolewski, Józef

    2017-01-01

    Target for HTR: Polish Heat Market: Today 100% heat market is dominated by fossil fuels; mostly coal in district heating and coal and gas in industry heat generation. Huge potential for nuclear reactors Currently can be addressed only in terms of LWR, i.e. T <250 ° C, useful in district heating, but not in industry. Need for new technologies •HTGR (High Temperature Gas Reactor) ~600°C, e.g. for industry steam generation. •VHTR (Very High Temperature Reactor), ... ~1000°C, e.g. for hydrogen production

  18. Relationship between the Toyo Tanso Group and HTR-PM

    International Nuclear Information System (INIS)

    Zhan Guobin; Konishi, Takashi

    2014-01-01

    IG-110 that is Isotropic graphite for nuclear applications, is the only product that is used for two types of High Temperature Gas-cooled Reactors, prismatic type HTTR and pebble-bed type HTR-10, that are currently in operation in the world. IG-110 is highly evaluated in the global market for its track record and physical stability. The Toyo Tanso Group won the contract to build graphite core internals for HTR-PM that is a world’s first modular pebble-bed high temperature gas-cooled demonstration reactor. A decision was made to manufacture IG-110 graphite materials at Toyo Tanso Japan called TTJ and to process products and undertake temporary assembly at Shanghai Toyo Tanso called STT. Manufacture of graphite materials for which TTJ is responsible has been completed. As the next step, processing of products is scheduled to commence at STT from this autumn. Our graphite materials were required to be 2,000 mm or more in maximum length. The number of graphite blocks required exceeded 3,500. Although the graphite structure requirements including configuration were highly challenging, we were able to meet all the requirements with our engineering capabilities, i.e. decades of track record in manufacture and stability in characteristics. STT that will start the machining process this autumn is equipped with state-of-the-art processing machines and three-dimensional measuring machines. Notably, STT has high levels of engineering capabilities to process and inspect tens of thousands of internal components for reactors in accordance with drawings and to temporarily assemble these components. (author)

  19. Brazing graphite to graphite

    International Nuclear Information System (INIS)

    Peterson, G.R.

    1976-01-01

    Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

  20. Dragon Drone UAV System

    Science.gov (United States)

    2003-09-02

    TYPE N/A 3. DATES COVERED - 4. TITLE AND SUBTITLE Dragon Drone UAV System 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER...ABSTRACT unclassified c. THIS PAGE unclassified Standard Form 298 (Rev. 8-98) Prescribed by ANSI Std Z39-18 A E R O S Y S T E M S BAI’s Dragon Drone ...the hundreds. BAI’s Dragon Drone system is the result of combining new ideas and emerging technologies with the in-depth knowl- edge gained from real

  1. HTR-10 severe accident management

    International Nuclear Information System (INIS)

    Xu Yuanhui; Sun Yuliang

    1997-01-01

    The High Temperature Gas-cooled Reactor (HTR-10) is under construction at the Institute of Nuclear Energy Technology site northwest of Beijing. This 10 MW thermal plant utilizes a pebble bed high temperature gas cooled reactor for a large range of applications such as electricity generation, steam and district heat generation, gas turbine and steam turbine combined cycle and process heat for methane reforming. The HTR-10 is the first high temperature gas cooled reactor to be licensed in China. This paper describes the safety characteristics and design criteria for the HTR-10 as well as the accident management and analysis required for the licensing process. (author)

  2. HTR-10 management information system

    International Nuclear Information System (INIS)

    Liu Ruoxiao; Wu Zhongwang; Xi Shuren

    2000-01-01

    The HTR-10 Management information system (REMIS) strengthens the managerial level and usage of the information of HTR-10, thereby enhances the ability and efficiency of the design and management work. REMIS is designed based on the Client/Server framework. Database management system is SQL Server 6.5 for NT, While the client side is developed by Borland C ++ Builder, and it is based on Windows 95/98. The network protocol is TCP/IP. REMIS collects date of the HTR-10 at four parameters: Reactor properties, Design parameters, Equipment properties Reactor system flow charts. Final discussing extended prospect of REMIS

  3. Special graphites; Graphites speciaux

    Energy Technology Data Exchange (ETDEWEB)

    Leveque, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

  4. A Wims E analysis of the zero energy experiments of Dragon prior to charge IV - Part 3: Prelmininary investigations into the control rod representation and related problems

    Energy Technology Data Exchange (ETDEWEB)

    Dubofsky, W; Woloch, F

    1973-09-28

    In the collaboration between UKAEA and Dragon for Zero Energy Experiment Evaluation, Dragon is to undertake all investigations needed for the representation of specific features of the Dragon Reactor experiment which are not necessarily characteristic of the large HTR system. One quite obvious uniqueness of Dragon is the location of the control rods situated in the radial reflector, only a few thermal mean free paths away from the core reflector boundary. In the first onslaught it is our intention to use the already available WIMS modules as much as possible and to identify difficulties in doing so. Finally it will be profitable to consider if the control rod representation found as a result of the whole exercise lends itself for the routine calculations of Dragon. In this paper Section 2 describes the geometrical difficulties to be overcome. Section 3 presents the results of calculations so far available. Section 4 gives the layout of further calculations planned and Section 5 presents the conclusions.

  5. The Research Status for Decommissioning and Radioactive Waste Minimization of HTR-PM

    International Nuclear Information System (INIS)

    Li Wenqian; Li Hong; Cao Jianzhu; Tong Jiejuan

    2014-01-01

    Decommissioning of the high-temperature gas-cooled reactor-pebble bed module (HTR-PM) as a part of the nuclear power plant, is very important during the early design stage of the construction, and it is under study and research currently. This article gives a thorough description of the current decommissioning study status of HTR-PM. Since HTR-PM has its features such as adopting a large amount of graphite, the waste inventory and characterization will be quite different from other type of reactors, new researches should be carried out and good lessons of practices and experiences should be learned from international other reactors, especially the AVR. Based on the new international regulations and Chinese laws, a comprehensive decommissioning program should be proposed to guarantee the HTR-PM will succeed in every stage of the decommissioning, such as defueling, decontamination, dismantling, demolition, waste classification and disposal, etc. In the meantime, the minimization of the radioactive waste should be taken into account during the whole process - before construction, during operation and after shut down. In this article, the decommissioning strategy and program conception of HTR-PM will be introduced, the radiation protection consideration during the decommissioning activities will be discussed, and the research on the activation problem of the decommissioning graphite will be introduced. (author)

  6. Oxidation kinetics of innovative carbon materials with respect to severe air ingress accidents in HTRs and graphite disposal or processing; Oxidationskinetik innovativer Kohlenstoffmaterialien hinsichtlich schwerer Lufteinbruchstoerfaelle in HTR's und Graphitentsorgung oder Aufbereitung

    Energy Technology Data Exchange (ETDEWEB)

    Schloegel, Baerbel

    2010-07-01

    Currently future nuclear reactor concepts of the Fourth Generation (Gen IV) are under development. To some extend they apply with new, innovative materials developed just for this purpose. This thesis work aims at a concept of Generation IV Very High Temperature Reactors (VHTR) in the framework of the European project RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation). The concept named ANTARES (AREVA New Technology based on advanced gas-cooled Reactors for Energy Supply) was developed by AEVA NP. It is a helium cooled, graphite moderated modular reactor for electricity and hydrogen production, by providing the necessary process heat due to its high working temperature. Particular attention is given here to oxidation kinetics of newly developed carbon materials (NBG-17) with still unknown but needed information in context of severe air ingress accident in VHTR's. Special interest is paid to the Boudouard reaction, the oxidation of carbon by CO{sub 2}. In case of an air ingress accident, carbon dioxide is produced in the primary reaction of atmospheric oxygen with reflector graphite. From there CO{sub 2} could flow into the reactor core causing further damage by conversion into CO. The purpose of this thesis is to ascertain if and to what degree this could happen. First of all oxidation kinetic data of the Boudouard reaction with NBG-17 is determined by experiments in a thermo gravimetric facility. The measurements are evaluated and converted into a common formula and a Langmuir-Hinshelwood similar oxidation kinetic equation, as input for the computer code REACT/THERMIX. This code is then applied to analyse severe air ingress accidents for several air flow rates. The results are discussed for two accident situations, in which a certain graphite burn off is achieved. All cases show much more damage to the graphite bottom reflector than to the reactor core. Thus the bottom reflector will lose its structural integrity much earlier than the

  7. Burner and dissolver off-gas treatment in HTR fuel reprocessing

    International Nuclear Information System (INIS)

    Barnert-Wiemer, H.; Heidendael, M.; Kirchner, H.; Merz, E.; Schroeder, G.; Vygen, H.

    1979-01-01

    In the reprocessing of HTR fuel, essentially all of the gaseous fission products are released during the heat-end tratment, which includes burning of the graphite matrix and dissolving of the heavy metallic residues in THOREX reagent. Three facilities for off-gas cleaning are described, the status of the facility development and test results are reported. Hot tests with a continuous dissolver for HTR-type fuel (throughput 2 kg HM/d) with a closed helium purge loop have been carried out. Preliminary results of these experiments are reported

  8. Review on characterization methods applied to HTR-fuel element components

    International Nuclear Information System (INIS)

    Koizlik, K.

    1976-02-01

    One of the difficulties which on the whole are of no special scientific interest, but which bear a lot of technical problems for the development and production of HTR fuel elements is the proper characterization of the element and its components. Consequently a lot of work has been done during the past years to develop characterization procedures for the fuel, the fuel kernel, the pyrocarbon for the coatings, the matrix and graphite and their components binder and filler. This paper tries to give a status report on characterization procedures which are applied to HTR fuel in KFA and cooperating institutions. (orig.) [de

  9. A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Marullo, G C

    1971-02-15

    In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

  10. Research on application of burnable poison in pebble bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhang Jian; Shan Wenzhi; Jing Xingqing

    2013-01-01

    Burnable poison in fuel ball was used in pebble bed high-temperature gas-cooled reactor (HTR) to optimize the shape and the peak factor of power distribution in certain conditions. Two options are available and evaluated, that is the homogeneous burnable poison in graphite matrix and burnable poison particles (BPPs) in fuel balls. Due to the absorption cross section of "1"0B, the depletion speed for homogeneous burnable poison is very fast, and difficult to control, on the other side, the depletion speed of BPPs can be optimized respecting to its size, and better shape and peak value of power distribution can be achieved. (authors)

  11. Proceedings of the workshop on structural design criteria for HTR

    International Nuclear Information System (INIS)

    Breitbach, G.; Schubert, F.; Nickel, H.

    1989-04-01

    The papers demonstrate the status of high temperature reactor technology with regard to its realization in the nuclear power industry of various countries (FRG, USA, Japan) as well as to the development of safety rules in Germany. The design criteria for HTR could be presented. The criteria already determine definitely and almost completely the relevant requirements of the component rules. The informations include the technical boundary conditions with regard to safety, the metallic high temperature components, a particular section dealing with the reactor pressure vessel, especially with the prestressed concrete vessel, and the structural graphite components. (DG)

  12. Bitten by a Dragon.

    Science.gov (United States)

    Ducey, Stephen D; Cooper, Jeffrey S; Wadman, Michael C

    2016-06-01

    Komodo dragons (Varanus komodoensis) are large lizards known to take down prey even larger than themselves. They rarely attack humans. A 38-year-old woman was bitten by a Komodo dragon on her hand while cleaning its enclosure. She was transiently hypotensive. The wounds were extensively cleaned, and she was started on prophylactic antibiotics. Her wounds healed without any infectious sequelae. Komodo dragon bites are historically thought to be highly infectious and venomous. Based on a literature review, neither of these are likely true. As in any bite, initial stabilization followed by wound management are the main components to therapy. Copyright © 2016 Wilderness Medical Society. Published by Elsevier Inc. All rights reserved.

  13. Nuclear astrophysics at DRAGON

    International Nuclear Information System (INIS)

    Hager, U.

    2014-01-01

    The DRAGON recoil separator is located at the ISAC facility at TRIUMF, Vancouver. It is designed to measure radiative alpha and proton capture reactions of astrophysical importance. Over the last years, the DRAGON collaboration has measured several reactions using both radioactive and high-intensity stable beams. For example, the 160(a, g) cross section was recently measured. The reaction plays a role in steady-state helium burning in massive stars, where it follows the 12C(a, g) reaction. At astrophysically relevant energies, the reaction proceeds exclusively via direct capture, resulting in a low rate. In this measurement, the unique capabilities of DRAGON enabled determination not only of the total reaction rates, but also of decay branching ratios. In addition, results from other recent measurements will be presented

  14. The Wave Dragon

    DEFF Research Database (Denmark)

    Sørensen, H. C.; Hansen, R.; Friis-Madsen, E.

    2000-01-01

    The Wave Dragon is an offshore wave energy converter of the overtopping type, utilizing a patented wave reflector design to focus the waves towards a ramp, and the overtopping is used for electricity production through a set of Kaplan/propeller hydro turbines. During the last 2 years, excessive...... design an testing has been performed on a scale 1:50 model of the Wave Dragon, and on a scale 1:3:5 model turbine. Thus survivability, overtopping, hydraulic response, turbine performance and feasibility have been verified....

  15. Analysis on First Criticality Benchmark Calculation of HTR-10 Core

    International Nuclear Information System (INIS)

    Zuhair; Ferhat-Aziz; As-Natio-Lasman

    2000-01-01

    HTR-10 is a graphite-moderated and helium-gas cooled pebble bed reactor with an average helium outlet temperature of 700 o C and thermal power of 10 MW. The first criticality benchmark problem of HTR-10 in this paper includes the loading number calculation of nuclear fuel in the form of UO 2 ball with U-235 enrichment of 17% for the first criticality under the helium atmosphere and core temperature of 20 o C, and the effective multiplication factor (k eff ) calculation of full core (5 m 3 ) under the helium atmosphere and various core temperatures. The group constants of fuel mixture, moderator and reflector materials were generated with WlMS/D4 using spherical model and 4 neutron energy group. The critical core height of 150.1 cm obtained from CITATION in 2-D R-Z reactor geometry exists in the calculation range of INET China, JAERI Japan and BATAN Indonesia, and OKBM Russia. The k eff calculation result of full core at various temperatures shows that the HTR-10 has negative temperature coefficient of reactivity. (author)

  16. The racing dragon

    CERN Multimedia

    2009-01-01

    Dating back nearly 2000 years, the ancient Chinese tradition of Dragon Boat Racing was originally a celebration that fell on the 5th day of the 5th lunar month as a gesture to please the Gods and bring forth necessary rains to cultivate the lands. Now the CERN Canoe and Kayak Club, too, participates in this tradition, though not so much to please the Gods on the ritualistic date, but to bring forth giant smiles on the faces of members. Dragon Boat Racing has been rising steadily in popularity in Europe since the mid nineties and with the great potential to host and promote Dragon Boat Racing in the Geneva area, the CERN Canoe and Kayak Club, has taken the initiative to bring the sport to the region. Some members of the Club traveled to Dole in June to participate in the Festival Dragon Boat 2009. Under perfect sunny conditions, the team triumphed in their first ever tournament, cruising to a convincing first place overall finish. T...

  17. Dragon Boat Festival.

    Science.gov (United States)

    Lew, Gordon

    This is one of a series of elementary readers written in Cantonese and English and designed to familiarize children with the traditional major Chinese festivals celebrated by the Chinese in America. This booklet describes the celebration of the Dragon Boat Festival, which marks the beginning of summer. A brief background to the festival is…

  18. Wave Dragon MW

    DEFF Research Database (Denmark)

    Kofoed, Jens Peter; Frigaard, Peter

    Wave Dragon is a wave energy converter of the overtopping type. The device has been thoroughly tested on a 1:51.8 scale model in wave laboratories and a 1:4.5 scale model deployed in Nissum Bredning, a large inland waterway in Denmark. Based on the experience gained a full scale, multi MW prototype...

  19. 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.)

  20. Characterization of graphite dust produced by pneumatic lift

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Kang, Feiyu [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Yang, Xiaoyong; Li, Weihua [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 100084 (China)

    2016-08-15

    Highlights: • Generation of graphite dust by pneumatic lift. • Determination of morphology and particle size distribution of graphite dust. • The size of graphite dust in this study is compared to AVR and THTR-300 results. • Graphite dust originates from both filler and binder of the matrix graphite. - Abstract: Graphite dust is an important safety concern of high-temperature gas-cooled reactor (HTR). The graphite dust could adsorb fission products, and the radioactive dust is transported by the coolant gas and deposited on the surface of the primary loop. The simulation of coagulation, aggregation, deposition, and resuspension behavior of graphite dust requires parameters such as particle size distribution and particle shape, but currently very limited data on graphite dust is available. The only data we have are from AVR and THTR-300, however, the AVR result is likely to be prejudiced by the oil ingress. In pebble-bed HTR, graphite dust is generally produced by mechanical abrasion, in particular, by the abrasion of graphite pebbles in the lifting pipe of the fuel handling system. Here we demonstrate the generation and characterization of graphite dust that were produced by pneumatic lift. This graphite dust could substitute the real dust in HTR for characterization. The dust, exhibiting a lamellar morphology, showed a number-weighted average particle size of 2.38 μm and a volume-weighted average size of 14.62 μm. These two sizes were larger than the AVR and THTR results. The discrepancy is possibly due to the irradiation effect and prejudice caused by the oil ingress accident. It is also confirmed by the Raman spectrum that both the filler particle and binder contribute to the dust generation.

  1. Worldwide status of HTR development

    International Nuclear Information System (INIS)

    1978-06-01

    The International Atomic Energy Agency convened a technical committee meeting on high temperature reactors (HTRs) from 12-14 Dec. 1977 at Agency Headquarters to provide a forum for the exchange of information on the status of HTR development programmes and to receive advice on the Agency programme in this field. The continuing high level of international interest in HTRs was evidenced by the participation from 11 countries and 2 organizations: Austria, Belgium, France, Federal Republic of Germany, Japan, Netherlands, Poland, Switzerland, Union of Soviet Socialist Republics, United Kingdom of Great Britain, United States of America, Commission of the European Communities, and the OECD Nuclear Energy Agency. In order to promote the continuing exchange of technical information through the offices of the IAEA, a recommendation was made that the Agency establish a standing International Working Group on High Temperature Reactors (IWGHTR). This recommendation is being implemented in 1978. Considerable information on recent progress in HTR development was present at the technical committee meeting in technical reports and in progress reports on HTR development programmes. Since this material will not be published, this summary report on the worldwide status of HTR development at the beginning of 1978 has been prepared, based primarily on information presented at the December 1977 meeting

  2. 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.)

  3. Design on Hygrometry System of Primary Coolant Circuit of HTR-PM

    International Nuclear Information System (INIS)

    Sun Yanfei; Zhong Shuoping; Huang Xiaojin

    2014-01-01

    Helium is the primary coolant in HTR-PM. If vapor get into the helium in primary coolant circuit because of some special reasons, such as the broken of steam-generator tube, chemical reaction will take effect between the graphite in reactor core and vapor in primary coolant circuit, and the safety of the reactor operation will be influenced. So the humidity of the helium in primary coolant circuit is one key parameter of HTR-PM to be monitored in-line. Once the humidity is too high, trigger signal of turning off the reactor must be issued. The hygrometry system of HTR-PM is consisting of filter, cooler, hygrometry sensor, flow meter, and some valves and tube. Helium with temperature of 250℃ is lead into the hygrometry system from the outlet of the main helium blower. After measuring, the helium is re-injected back to the primary circuit. No helium loses in this processing, and no other pump is needed. Key factors and calculations in design on hygrometry system of HTR-PM are described. A sample instrument has been made. Results of experiments proves that this hygrometry system is suitable for monitoring the humidity of the primary coolant of HTR-PM. (author)

  4. Oxidation of carbon based material for innovative energy systems (HTR, fusion reactor): status and further needs

    International Nuclear Information System (INIS)

    Moormann, R.; Hinssen, H.K.; Latge, Ch.; Dumesnil, J.; Veltkamp, A.C.; Grabon, V.; Beech, D.; Buckthorpe, D.; Dominguez, T.; Krussenberg, A.K.; Wu, C.H.

    2000-01-01

    Following an overview on kinetics of carbon/gas reactions, status and further needs in selected safety relevant fields of graphite oxidation in high temperature reactors (HTRs) and fusion reactors are outlined. Kinetics was detected due to the presence of such elements as severe air ingress, lack of experimental data on Boudouard reaction and a similar lack of data in the field of advanced oxidation. The development of coatings which protect against oxidation should focus on stability under neutron irradiation and on the general feasibility of coatings on HTR pebble fuel graphite. Oxidation under normal operation of direct cycle HTR requires examinations of gas atmospheres and of catalytic effects. Advanced carbon materials like CFCs and mixed materials should be developed and tested with respect to their oxidation resistance in a common HTR/fusion task. In an interim HTR, fuel storage radiolytic oxidation under normal operation and thermal oxidation in accidents have to be considered. Plans for future work in these fields are described. (authors)

  5. Parthenogenesis in Komodo dragons.

    Science.gov (United States)

    Watts, Phillip C; Buley, Kevin R; Sanderson, Stephanie; Boardman, Wayne; Ciofi, Claudio; Gibson, Richard

    2006-12-21

    Parthenogenesis, the production of offspring without fertilization by a male, is rare in vertebrate species, which usually reproduce after fusion of male and female gametes. Here we use genetic fingerprinting to identify parthenogenetic offspring produced by two female Komodo dragons (Varanus komodoensis) that had been kept at separate institutions and isolated from males; one of these females subsequently produced additional offspring sexually. This reproductive plasticity indicates that female Komodo dragons may switch between asexual and sexual reproduction, depending on the availability of a mate--a finding that has implications for the breeding of this threatened species in captivity. Most zoos keep only females, with males being moved between zoos for mating, but perhaps they should be kept together to avoid triggering parthenogenesis and thereby decreasing genetic diversity.

  6. Here be no dragons

    International Nuclear Information System (INIS)

    1987-01-01

    ''Here be dragons'' is the phase used by ancient map makers to indicate areas about which they knew nothing or which they suspected contained unknown dangers. The aim of this booklet, ''Here be no dragons'', is to dispel the myths, misconceptions and misinformation about nuclear power. The South of Scotland Electricity Board explains why nuclear power is important to Scotland economically and deals in a non-technical way with many of the safety issues raised by the presence and operation of nuclear reactors. The environmental issues are also presented simply, with an explanation of the average annual radiation dose to the population of the UK, and a comparison of the radiation doses from the Chernobyl accident, compared to variations in background doses. The risks from nuclear accidents and the risk of death from accidents in industries in the UK compared with the risk from cancers potentially produced among radiation workers, are compared. (U.K.)

  7. Dragon Lake, Siberia

    Science.gov (United States)

    2002-01-01

    Nicknamed 'Dragon Lake,' this body of water is formed by the Bratskove Reservoir, built along the Angara river in southern Siberia, near the city of Bratsk. This image was acquired in winter, when the lake is frozen. This image was acquired by Landsat 7's Enhanced Thematic Mapper plus (ETM+) sensor on December 19, 1999. This is a natural color composite image made using blue, green, and red wavelengths. Image provided by the USGS EROS Data Center Satellite Systems Branch

  8. Graphite for high-temperature reactors

    International Nuclear Information System (INIS)

    Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

    1976-01-01

    The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

  9. Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Tang Chunhe

    2014-01-01

    Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

  10. Wave Dragon Buoyancy Regulation Study

    DEFF Research Database (Denmark)

    Jakobsen, Jens; Kofoed, Jens Peter

    Wave Dragon is a wave energy converter, which was deployed offshore at Nissum Bredning in Denmark in 2003. The experience gained from operating Wave Dragon during 2003 and 2004 has shown that the buoyancy regulation system can be improved in a number of ways. This study describes the current...

  11. The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

    Energy Technology Data Exchange (ETDEWEB)

    Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

    2017-03-15

    Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

  12. The challenge of introducing HTR plants on to the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    The international power plant market today is characterized by high increase in energy consumption for developing countries with limitations of investment capital and low increase in energy consumption for industrialized countries with limitations of additional power plant capacities. As a consequence there is a low demand for large new power stations. This leads to a tendency for small and medium sized power plant units - meeting high environmental standards - for which the total investment volume is low and full load operation of a plant can be realized earlier due to the small block capacity. - For nuclear power plants the High-Temperature-Reactor (HTR)-line with spherical fuel elements and a core structure of graphite is specially suited for this small and medium sized nuclear reactor (SMSNR) capacity. The excellent safety characteristics, the high availability, the low radiation doses for the operation personnel and the environment of the HTR line has been demonstrated by 20 years of operation of the AVR-15 MWe experimental power plant in Juelich F.R.G. and since 1985 by operation of the THTR-300 MWe prototype plant at Hamm-Uentrop F.R.G. Up-dated concepts of the HTR-line are under design for electricity generation (HTR-500), for co-generation of power and heat (HTR-100) and for district heating purposes only (GHR-10). By implementing two HTR projects the Brown Boveri Group is in the position to realize the collected experiences from design, licensing, erection, commissioning and operation for the follow-on projects. This leads to practical and sound technical solutions convenient for existing manufacturing processes, well known materials, standardized components and usual manufacturing tolerances. Specific plant characteristics can be used for advantages in the competition. (author)

  13. Progress of the HTR-10 project

    International Nuclear Information System (INIS)

    Zhong, D.; Xu, Y.

    1996-01-01

    This paper briefly introduces the main technical features and the design specifications of the HTR-10. Present status and main progress of the license applications, the design and manufacture of the main components and the engineering experiments as well as the construction of the HTR-10 are summarized. (author). 3 tabs

  14. Special graphites

    International Nuclear Information System (INIS)

    Leveque, P.

    1964-01-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

  15. The Dragon Bone Collectors

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Finding of a dinosaur fossil skeleton sparks excitement among paleontologists and locals in a poor Henan village Avillager’s accidental discovery four years ago has made known to the world a rich mine of dinosaur fossils in Ruyang County,central China’s Henan Province.At the same time,the fate of the small village has been changed. Li Chui,a farmer in Shaping Village, thought he had found bones of a"dragon"when he dug up stones for his new house on an April morning in 2005.

  16. Costs of head-end incineration with respect to Kr separation in the reprocessing of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Barnert-Wiemer, H.; Boehnert, R.

    1976-07-15

    The C-incinerations and the Kr-separations during head-end incineration in the reprocessing of HTR fuel elements are described. The costs for constructing an operating a head-end incineration of reprocessing capacities with 5,000 to 50,000 MW(e)-HTR power have been determined. The cost estimates are divided into investment and operating costs, further after the fraction of the N/sub 2/-content in the incineration exhaust gas, which strongly affects costs. It appears that, in the case of Kr-separation from the incineration exhaust gas, the investment costs as well as the operating costs of the head-end for N/sub 2/-containing exhaust gas are considerably greater than those for gas without N/sub 2/. The C-incineration of the graphite of the HTR fuel elements should therefore only be performed with influx gas that is free of N/sub 2/.

  17. The Hitrex Programme: unperturbed HTR lattice and control rod measurements

    Energy Technology Data Exchange (ETDEWEB)

    Beynon, A J; Nunn, D L

    1972-06-15

    Reactivity, power distributions, plutonium production and fast neutron graphite damage are being studied at Berkeley Nuclear Laboratories (BNL) on the HTR 'Hitrex' reactor under cold clean conditions. Rod interactions, important in assessing local criticality hazards, are receiving special attention in the measurements. The proposals for the first two series of measurements on Hitrex are discussed in this note, Hitrex 1a being the unperturbed reactor, and Hitrex 1b the same fuel array but with a number of different control absorber loadings in it. Common to both series will be cross pin, cross block and cross core measurements of power rating, thermal spectrum and damage dose distributions, so that these will be known as functions of the fuel, reflector and absorber environment.

  18. Status of the HTR 500 design program

    International Nuclear Information System (INIS)

    Baust, E.; Arndt, E.

    1987-01-01

    Since 1982 BBC/HRB have offered the HTR 500 as the follow-on project of the THTR 300, the first large pebble bed reactor. The technical concept of the HTR-500 largely corresponds to the THTR 300 which has been in operation for almost 2 years now. In developing the design concept of the HTR 500 the ideas and demands of the reactor users in the FRG interested in the HTR have been taken into consideration to a large extent. In 1982 these potential users formed a working group 'Arbeitsgemeinschaft Hochtemperaturreaktor' (AHR), representing 16 power indusry companies and in early 1983, awarded a contract to HRB to perform a conceptual design study on the HTR 500. Within this conceptual design study BBC/HRB developed the safety concept of the HTR 500, prepared a detailed description of the overall power plant, and performed a cost calculation. These activities were completed in 1984. Based on the positive results of this conceptual design study, BBC/HRB are expecting to be granted a design contract by the users company Hochtemperaturreaktor GmbH (HRG) to establish the final complete design plans and documents for the HTR 500. (author)

  19. Prediction calculation of HTR-10 fuel loading for the first criticality

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Gu Yuxiang; Shan Wenzhi

    2001-01-01

    The 10 MW high temperature gas cooled reactor (HTR-10) was built at Institute of Nuclear Energy Technology, Tsinghua University, and the first criticality was attained in Dec. 2000. The high temperature gas cooled reactor physics simulation code VSOP was used for the prediction of the fuel loading for HTR-10 first criticality. The number of fuel element and graphite element was predicted to provide reference for the first criticality experiment. The prediction calculations toke into account the factors including the double heterogeneity of the fuel element, buckling feedback for the spectrum calculation, the effect of the mixture of the graphite and the fuel element, and the correction of the diffusion coefficients near the upper cavity based on the transport theory. The effects of impurities in the fuel and the graphite element in the core and those in the reflector graphite on the reactivity of the reactor were considered in detail. The first criticality experiment showed that the predicted values and the experiment results were in good agreement with little relative error less than 1%, which means the prediction was successful

  20. Victorian Dragons: The Reluctant Brood.

    Science.gov (United States)

    Berman, Ruth

    1984-01-01

    Relates why nineteenth century fantasy writers shied away from the use of dragons in their stories and rejoices over the return and happy transformation of this mythical beast in children's literature. (HOD)

  1. Impact of Wave Dragon on Wave Climate

    DEFF Research Database (Denmark)

    Andersen, Thomas Lykke; Tedd, James; Kramer, Morten

    This report is an advisory paper for use in determining the wave dragon effects on hydrography, by considering the effect on the wave climate in the region of a wave dragon. This is to be used in the impact assessment for the Wave Dragon pre-commercial demonstrator.......This report is an advisory paper for use in determining the wave dragon effects on hydrography, by considering the effect on the wave climate in the region of a wave dragon. This is to be used in the impact assessment for the Wave Dragon pre-commercial demonstrator....

  2. Predictions of the Bypass Flows in the HTR-PM Reactor Core

    International Nuclear Information System (INIS)

    Sun Jun; Chen Zhipeng; Zheng Yanhua; Shi Lei; Li Fu

    2014-01-01

    In the HTR-PM reactor core, the basic structure materials are large amount of graphite reflectors and carbon bricks. Small gaps among those graphite and carbon bricks are widespread in the reactor core so that the cold helium flow may be bypassed and not completely heated. The bypass flows in relative lower temperature would change the flow and temperature distributions in the reactor core, therefore, the accurate prediction of bypass flows need to be carried out carefully to evaluate the influence to the reactor safety. Based on the characteristics of the bypass flow problem, hybrid method of the flow network and the CFD tools was employed to represent the connections and calculate flow distributions of all the main flow and bypass flow paths. In this paper, the hybrid method was described and applied to specific bypass flow problem in the HTR-PM. Various bypass flow paths in the HTR-PM were reviewed, figured out, and modeled by the flow network and the CFD methods, including the axial vertical gaps in the side reflectors, control rod channels, absorber sphere channels and radial gap flow through keys around the hot helium plenum. The bypass flow distributions and its flow rate ratio to the total flow rate in the primary loop were also calculated, discussed and evaluated. (author)

  3. VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

    International Nuclear Information System (INIS)

    Boehnert, R.; Walter, C.

    The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given

  4. VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

    Energy Technology Data Exchange (ETDEWEB)

    Boehnert, R.; Walter, C.

    1977-02-15

    The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given. (LK)

  5. Neutronic feasibility design of a small long-life HTR

    International Nuclear Information System (INIS)

    Ding Ming; Kloosterman, Jan Leen

    2011-01-01

    Highlights: ► We propose the neutronic feasibility design of a small, long lifetime and transportable HTR. ► Comparison of cylindrical, annular and scatter cores of the small block-type HTR. ► The design of the scatter core effectively reduces the number of the fuel block and increases the lifetime and burnup of the reactor. - Abstract: Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U

  6. An analysis of irradiation creep in nuclear graphites

    International Nuclear Information System (INIS)

    Neighbour, G.B.; Hacker, P.J.

    2002-01-01

    Nuclear graphite under load shows remarkably high creep ductility with neutron irradiation, well in excess of any strain experienced in un-irradiated graphite (and additional to any dimensional changes that would occur without stress). As this behaviour compensates, to some extent, some other irradiation effects such as thermal shutdown stresses, it is an important property. This paper briefly reviews the approach to irradiation creep in the UK, described by the UK Creep Law. It then offers an alternative analysis of irradiation creep applicable to most situations, including HTR systems, using AGR moderator graphite as an example, to high values of neutron fluence, applied stress and radiolytic weight loss. (authors)

  7. Raw materials for reflector graphite (for reactors)

    International Nuclear Information System (INIS)

    Wilhelmi, G.; Mindermann, D.

    1992-01-01

    The manufacturing concept for the core components of German high temperature reactor (HTR) types of graphite was previously entirely directed to the use of German tar coke (St coke). As the plants for producing this material no longer complied technically with the current environmental protection requirements, one had to assume that they would soon be shut down. To prevent bottlenecks in the erection of future HTR plants, alternative cokes produced by modern processes by Japanese manufacturers were checked for their suitability for the manufacture of reactor graphite. This report describes the investigations carried out on these materials from the safe delayed coking process. The project work, apart from analysis of the main data of the candidate coke considered, included the processing of the raw materials into directly and secondarily extruded graphite rods on the laboratory scale, including characterisation. As the results show, the material data achieved with the previous raw material can be reproduced with Japanese St coke. The tar coke LPC-A from the Nippon Steel Chemical Co., Ltd was decided on as the new standard coke for manufacturing reflector graphite. (orig.) With 15 tabs., 2 figs [de

  8. Structural strength of core graphite bars

    International Nuclear Information System (INIS)

    Kikuchi, K.; Futakawa, M.

    1987-01-01

    A HTR core consists of fuel, hot plenum, reflector and thermal barrier blocks. Each graphite block is supported by three thin cylindrical graphite bars called support post. Static and dynamic core loads are transmitted by the support posts to the thermal barrier blocks and a support plate. These posts are in contact with the blocks through hemispherical post seats to absorb the relative displacement caused by seismic force and the difference of thermal expansion of materials at the time of the start-up and shutdown of a reactor. The mixed fracture criterion of principal stress and modified Mohr-Coulomb's theory as well as the fracture criterion of principal stress based on elastic stress analysis was discussed in connection with the application to HTR graphite components. The buckling fracture of a support post was taken in consideration as one of the fracture modes. The effect that the length/diameter ratio of a post, small rotation and the curvature of post ends and seats exerted on the fracture strength was studied by using IG-110 graphite. Contacting stress analysis was carried out by using the structural analysis code 'COSMOS-7'. The experimental method, the analysis of buckling strength and the results are reported. The fracture of a support post is caused by the mixed mode of bending deformation, split fracture and shearing fracture. (Kako, I.)

  9. For a Global HTR Marketing Initiative

    International Nuclear Information System (INIS)

    Bredimas, Alexandre; Venneri, Francesco; Richards, Matthew

    2014-01-01

    HTRs are at a crossroads in their history. The technology is proven and the current technical developments relatively mastered but the marketing track record is disappointing. This paper comes to the conclusion that an international, collaborative marketing and communication plan must be implemented in order to address the marketing bottleneck of HTRs. The paper reflects about the HTR product specificities, its unique selling points and its positioning against other nuclear designs and gas cogeneration. It summarises the global market status and demonstrates that the global market for HTRs is there, for electricity generation, industrial cogeneration and polygeneration. The paper finally argues that HTR vendors have a shared interest to unite in order to succeed in activating the market demand for HTR, and suggests an action plan for an international collaboration among HTR vendors to market and communicate globally on HTRs and reach together a critical mass of business leads worldwide, a mutually beneficial outcome. (author)

  10. Gamma scanning of full scale HTR fuel elements

    International Nuclear Information System (INIS)

    Harrison, T.A.; Simpson, J.A.H.; Nabielek, H.

    1983-04-01

    Gamma scanning for the determination of burn-up and fission product inventory has been developed at the Dragon Project, suitable for measurements on fuel elements and segments from full-sized integral block elements. This involved the design and construction of a new lead flask with sophisticated collimator design. State-of-the art gamma spectrometric equipment was set up to cope with strong variations of count-rate and high data throughput. Software efforts concentrated on the calculation of the self absorption and absorption corrections in the complicated geometry of multi-hole graphite block segments with a corrugated circumference. The techniques described here are applicable to the non-destructive examination of a wide range of fuel element designs. (author)

  11. Notes on HTR applications in methanol production

    International Nuclear Information System (INIS)

    Santoso, B.; Barnert, H.

    1997-01-01

    Notes on the study of HTR applications are presented. The study in particular should be directed toward the most feasible applications of HTR for process heat generation. A prospective study is the conversion of CO 2 gas from Natuna to methanol or formic acid. Further studies needs to be deepened under the auspices of IAEA and countries that have similar interest. (author). 3 refs, 3 figs

  12. KWU's modular approach to HTR commercialization

    International Nuclear Information System (INIS)

    Frewer, H.; Weisbrodt, I.

    1983-01-01

    As a way of avoiding the uncertainties, delays and unacceptable commercial risks which have plagued advanced reactor projects in Germany, KWU is advocating a modular approach to commercialization of the high-temperature reactor (HTR), using small size standard reactor units. KWU has received a contract for the study of a co-generation plant based on this modular system. Features of the KWU modular HTR, process heat, gasification, costs and future development are discussed. (UK)

  13. Injuries in Competitive Dragon Boating.

    Science.gov (United States)

    Mukherjee, Swarup; Leong, Hin Fong; Chen, Simin; Foo, Yong Xiang Wayne; Pek, Hong Kiat

    2014-11-01

    Dragon boating is a fast-growing team water sport and involves forceful repetitive motions that predispose athletes to overuse injuries. Despite the rising popularity of the sport, there is a lack of studies on injury epidemiology in dragon boating. To investigate the injury epidemiology in competitive dragon boating athletes. Descriptive epidemiological study. A total of 95 dragon boaters (49 males, 46 females) representing their respective universities took part in this study. Data were collected retrospectively using a reliable and valid self-report questionnaire. The study period was from August 2012 to July 2013. A total of 104 musculoskeletal injuries were reported (3.82 injuries/1000 athlete-exposures), 99% of which occurred during training. The most commonly injured regions were the lower back (22.1%), shoulder (21.1%), and wrist (17.3%). The majority of injuries were due to overuse (56.3%), and incomplete muscle-tendon strain was the most prevalent type of injury (50.5%). The time loss from injuries varied. In addition, a significant majority of the dragon boating athletes incurred nonmusculoskeletal injuries, with abrasions (90.5%), blisters (78.9%), and sunburns (72.6%) being the most common. Competitive dragon boating has a moderately high injury incidence, and there seems to be a direct relationship between exposure time and injury rate. A majority of the injuries are overuse in nature, and the body parts most actively involved in paddling movement are at higher risk of injuries. The high incidence of nonmusculoskeletal injuries in dragon boaters suggested that these injuries are likely outcomes of participation in the sport.

  14. Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

    Science.gov (United States)

    Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

  15. The Renewal of HTR Development in Europe

    International Nuclear Information System (INIS)

    Hittner, Dominique

    2002-01-01

    The European HTR-Technology Network (HTR-TN), created in 2000, presently groups 20 organisations from European nuclear research and industry for developing the technologies of direct-cycle modular HTRs, which presently raise a large world-wide interest, because of their high potential for economic competitiveness, natural resource sparing, safety and minimisation of the waste impacts, in line with the goals of sustainable development of Generation IV. All aspects of HTR technologies are addressed by HTR-TN, from the reactor physics to the development of materials, fuel and components. Most of this activity is supported by the European Commission in the frame of its 5. EURATOM Framework Programme. The first results of HTR-TN programme are given: the analysis of the reactor physics international benchmark on the commissioning tests of HTTR (Japan), the long term behaviour of spent HTR fuel in geologic disposal conditions, the preparation of a very high burnup fuel irradiation and the development of fabrication processes for producing high performance coated particles, etc. (authors)

  16. Predictions on an HTR coolant composition after operational experience with experimental reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1981-01-01

    Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

  17. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  18. HTR combustion head end comparison of the shaft furnace and fluidized bed processes

    Energy Technology Data Exchange (ETDEWEB)

    Boehnert, R.; Kaiser, G.; Pirk, H.; Tillessen, U.

    1975-01-15

    Two methods are described for the combustion of the graphite of HTR fuel elements, a sufficient description of the principles being given to permit an understanding of the processes. The present state of the technology of the two processes is then compared on the basis of the results obtained at Gulf General Atomic. Finally, the possibilities of further development are examined using a pilot plant designed to deliver a reactor power of 7000 MWe as the basis. The present report is a collection of facts. It contains neither an evaluation nor a recommendation. A summarized comparison of the state of the technology and the possibilities of development is given in tabular form.

  19. The HTR-PM Plant Full Scope Training Simulator

    International Nuclear Information System (INIS)

    Wang Junsan; Wang Yuding; Zhou Shuyong; Cai Ruizhong; Cao Jianting

    2014-01-01

    This paper describes the technical aspects of the Full Scope Training Simulator developed for HTR-PM Plant in Shidao Bay, Shandong Province, China. An overview of the HTR-PM plant and simulator structure is presented. The models developed for the simulator are discussed in detail. Some important verification tests have been conducted on the HTR-PM Plant Training Simulator. (author)

  20. The HTR safety concept demonstrated by selected examples

    International Nuclear Information System (INIS)

    Sommer, H.; Stoelzl, D.

    1981-01-01

    The licensing experience gained in the Federal Republic of Germany is based on the licensing procedures for the THTR-300 and the HTR-1160. In the course of the licensing procedures for these reactors a safety concept for an HTR has been developed. This experience constitutes the basis for the design of future HTR's. (author)

  1. HTR core physics analysis at NRG

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

    2002-01-01

    Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

  2. Burning minor actinides in a HTR energy spectrum

    International Nuclear Information System (INIS)

    Pohl, Christoph; Rütten, H. Jochem

    2012-01-01

    Highlights: ► Burn-up analysis for varying plutonium/minor actinide fuel compositions. ► The influence of varying heavy metal fuel element loads is investigated. ► Significant burn-up via radiative capture and subsequently fission is observed. ► Difference observed between fuel element burn-up and total actinide burning rate. - Abstract: The generation of nuclear energy by means of the existing nuclear reactor systems is based mainly on the fission of U-235. But this comes along with the capture of neutrons by the U-238 faction and results in a build-up of plutonium isotopes and minor actinides as neptunium, americium and curium. These actinides are dominant for the long time assessment of the radiological risk of a final disposal therefore a minimization of the long living isotopes is aspired. Burning the actinides in a high temperature helium cooled graphite moderated reactor (HTR) is one of these options. The use of plutonium isotopes to sustain the criticality of the system is intended to avoid on the one hand highly enriched uranium because of international regulations and on the other hand low enriched uranium because of the build up of new actinides from neutron capture in the U-238 fraction. Because initial minor actinide isotopes are typically not fissionable by thermal neutrons the idea is to fission instead the intermediate isotopes generated by the first neutron capture. This paper comprises calculations for plutonium/minor actinides/thorium fuel compositions and their correlated final burn-up for a generic pebble bed HTR based on the reference design of the 400 MW PBMR. In particular the cross sections and the neutron balance of the different minor actinide isotopes in the higher thermal energy spectrum of a HTR will be discussed. For a fuel mixture of plutonium and minor actinides a significant burn-up of these actinides up to 20% can be achieved but at the expense of a higher residual fraction of plutonium in the burned fuel. Combining

  3. MCNP qualification on the HTR critical configurations: HTTR, HTR10 and PROTEUS results

    Energy Technology Data Exchange (ETDEWEB)

    TRAKAS, Christos; STOVEN, Gilles [AREVA NP, Tour Areva, 92084 Paris La Defence Cedex (France)

    2008-07-01

    Recent critical experiments, including PROTEUS, HTTR and HTR-10 provide a reliable qualification base for HTR criticality predictions. The fuel tested in these experiments, be it hexagonal block or pebble type, is irradiated in a spectrum comparable to that of the HTR planned by AREVA NP. The neutron spectrum is comparable in all three cases; the mean C/M value for all critical cases is less than +350 pcm (JEF2.2), +250 pcm (JEFF3.1) and +60 pcm (ENDF BVI). The C/M obtained for the rods worth, the reaction rates and the isothermal coefficient are very satisfactory. (authors)

  4. Preliminary HECTOR analysis by Dragon

    Energy Technology Data Exchange (ETDEWEB)

    Presser, W; Woloch, F

    1972-06-02

    From the different cores measured in HECTOR, only ACH 4/B-B was selected for the Dragon analysis, since it presented the largest amount of uniform fuel loading in the central test region and is therefore nearest to an infinite lattice. Preliminary results are discussed.

  5. Towards a more plausible dragon

    Science.gov (United States)

    Efthimiou, Costas

    2014-08-01

    Wizards, mermaids, dragons and aliens. Walking, running, flying and space travel. A hi-tech elevator, a computer, a propulsion engine and a black hole. What do all of these things have in common? This might seem like a really hard brainteaser but the answer is simple: they all obey the fundamental laws of our universe.

  6. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  7. Capital costs of modular HTR reactors

    International Nuclear Information System (INIS)

    Kugeler, K.; Froehling, W.

    1993-01-01

    A decisive factor in the introduction of a reactor line, in addition of its safety, which should exclude releases of radioactivity into the environment, is its economic development and, consequently, its competitiveness. The costs of the pressurized water reactor are used for comparison with the modular HTR reactor. If the measures proposed for evolutionary increases in safety of the PWR are taken, cost increases will have to be expected for that line. The modular HTR can now attain specific construction costs of 3000 deutschmarks per electric kilowatt. Mass production and the introduction of cost-reducing innovations can improve the economy of this line even further. In this way, the modular HTR concept offers the possibility to vendors and operators to set up new economic yardsticks in safety technology. (orig.) [de

  8. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  9. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Bucher, K.H.; Chawla, R.; Foskolos, K.; Luchsinger, H.; Mathews, D.; Sarlos, G.; Seiler, R.

    1990-01-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  10. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R; Bucher, K H; Chawla, R; Foskolos, K; Luchsinger, H; Mathews, D; Sarlos, G; Seiler, R [Paul Scherrer Institute, Laboratory for Reactor Physics and System Technology Wuerenlingen and Villigen, Villigen PSI (Switzerland)

    1990-07-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  11. HTR fuel development for advanced application

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.; Graham, L.W.; Hick, H.

    1975-01-01

    The advantages of the HTR for nuclear steam supply systems are briefly outlined. Due to its great design flexibility a number of different designs have evolved and the main characteristics of existing experimental prototype and power reactor HTR designs are summarized. The present state of coated particle fuel, particularly with regard to performance, is considered. Some implications of producing higher temperatures are discussed. Finally some of the developments in progress such as minimising the temperature drop between fuel and coolant, and of improving fuel performance by better fission product retention, better chemical stability, and the use of alternative coated materials, are discussed. (U.K.)

  12. Actinide production in different HTR-fuel cycle concepts

    International Nuclear Information System (INIS)

    Filges, D.; Hecker, R.; Mirza, N.; Rueckert, M.

    1978-01-01

    At the 'Institut fuer Reaktorentwicklung der Kernforschungsanlage Juelich' the production of α-activities in the following HTR-OTTO cycle concepts were studied: 1. standard HTR cycle (U-Th); 2. low enriched HTR cycle (U-Pu); 3. near breeder HTR cycle (U-Th); 4. combined system (conventional and near breeder HTR). The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The production of α-activity in HTR Uranium-Thorium fuel cycles has been investigated and compared with the standard LWR cycles. The calculations were performed by the short depletion code KASCO and the well-known ORIGEN program

  13. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K.; Cho, M. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW{sub th} block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW{sub th} block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared.

  14. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    International Nuclear Information System (INIS)

    Kim, Young Min; Jo, C. K.; Cho, M. S.

    2016-01-01

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW_t_h block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW_t_h block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared

  15. US/FRG umbrella agreement for cooperation in GCR Development. Fuel, fission products, and graphite subprogram. Quarterly status report, July 1, 1982-September 30, 1982

    International Nuclear Information System (INIS)

    Turner, R.F.

    1982-10-01

    This report describes the status of the cooperative work being performed in the Fuel, Fission Product, and Graphite Subprogram under the HTR-Implementing Agreement of the United States/Federal Republic of Germany Umbrella Agreement for Cooperation in GCR Development. The status is described relative to the commitments in the Subprogram Plan for Fuel, Fission Products, and Graphite, Revision 5, April 1982. The work described was performed during the period July 1, 1982 through September 30, 1982 in the HTGR Base Technology Program at Oak Ridge National Laboratory, the HTGR Fuel and Plant Technology Programs at General Atomic Company (GA), and the Project HTR-Brennstoffkreislauf of the Entwicklungsgemeinschaft HTR at KFA Julich, HRB Mannheim, HOBEG Hanau, and SIGRI Meitingen. The requirement for and format of this quarterly status report are specified in the HTR Implementing Agreement procedures for cooperation. Responsibility for preparation of the quarterly report alternates between GA and KFA

  16. Artificial graphites

    International Nuclear Information System (INIS)

    Maire, J.

    1984-01-01

    Artificial graphites are obtained by agglomeration of carbon powders with an organic binder, then by carbonisation at 1000 0 C and graphitization at 2800 0 C. After description of the processes and products, we show how the properties of the various materials lead to the various uses. Using graphite enables us to solve some problems, but it is not sufficient to satisfy all the need of the application. New carbonaceous material open application range. Finally, if some products are becoming obsolete, other ones are being developed in new applications [fr

  17. SpaceX Dragon Air Circulation System

    Science.gov (United States)

    Hernandez, Brenda; Piatrovich, Siarhei; Prina, Mauro

    2011-01-01

    The Dragon capsule is a reusable vehicle being developed by Space Exploration Technologies (SpaceX) that will provide commercial cargo transportation to the International Space Station (ISS). Dragon is designed to be a habitable module while it is berthed to ISS. As such, the Dragon Environmental Control System (ECS) consists of pressure control and pressure equalization, air sampling, fire detection, illumination, and an air circulation system. The air circulation system prevents pockets of stagnant air in Dragon that can be hazardous to the ISS crew. In addition, through the inter-module duct, the air circulation system provides fresh air from ISS into Dragon. To utilize the maximum volume of Dragon for cargo packaging, the Dragon ECS air circulation system is designed around cargo rack optimization. At the same time, the air circulation system is designed to meet the National Aeronautics Space Administration (NASA) inter-module and intra-module ventilation requirements and acoustic requirements. A flight like configuration of the Dragon capsule including the air circulation system was recently assembled for testing to assess the design for inter-module and intra-module ventilation and acoustics. The testing included the Dragon capsule, and flight configuration in the pressure section with cargo racks, lockers, all of the air circulation components, and acoustic treatment. The air circulation test was also used to verify the Computational Fluid Dynamics (CFD) model of the Dragon capsule. The CFD model included the same Dragon internal geometry that was assembled for the test. This paper will describe the Dragon air circulation system design which has been verified by testing the system and with CFD analysis.

  18. HTR fuel research in the HTR-TN network on the high flux reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M. [HFR Unit, Institute for Advanced Materials, European Commission, Joint Research Centre, Petten (Netherlands); Languille, A. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Guillermier, P. [FRAMATOME ANP, 69 - Lyon (France); Bakker, K. [Nuclear Research and Consultancy Group, Petten (Netherlands); Nabielek, H. [Forschungszentrum Juelich (Germany)

    2001-07-01

    Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

  19. HTR fuel research in the HTR-TN network on the high flux reactor

    International Nuclear Information System (INIS)

    Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M.; Languille, A.; Guillermier, P.; Bakker, K.; Nabielek, H.

    2001-01-01

    Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

  20. Red Dragon drill missions to Mars

    Science.gov (United States)

    Heldmann, Jennifer L.; Stoker, Carol R.; Gonzales, Andrew; McKay, Christopher P.; Davila, Alfonso; Glass, Brian J.; Lemke, Larry L.; Paulsen, Gale; Willson, David; Zacny, Kris

    2017-12-01

    We present the concept of using a variant of a Space Exploration Technologies Corporation (SpaceX) Dragon space capsule as a low-cost, large-capacity, near-term, Mars lander (dubbed ;Red Dragon;) for scientific and human precursor missions. SpaceX initially designed the Dragon capsule for flight near Earth, and Dragon has successfully flown many times to low-Earth orbit (LEO) and successfully returned the Dragon spacecraft to Earth. Here we present capsule hardware modifications that are required to enable flight to Mars and operations on the martian surface. We discuss the use of the Dragon system to support NASA Discovery class missions to Mars and focus in particular on Dragon's applications for drilling missions. We find that a Red Dragon platform is well suited for missions capable of drilling deeper on Mars (at least 2 m) than has been accomplished to date due to its ability to land in a powered controlled mode, accommodate a long drill string, and provide payload space for sample processing and analysis. We show that a Red Dragon drill lander could conduct surface missions at three possible targets including the ice-cemented ground at the Phoenix landing site (68 °N), the subsurface ice discovered near the Viking 2 (49 °N) site by fresh impact craters, and the dark sedimentary subsurface material at the Curiosity site (4.5 °S).

  1. Reactor physics calculations on HTR type configurations

    Energy Technology Data Exchange (ETDEWEB)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).

  2. Reactor physics calculations on HTR type configurations

    International Nuclear Information System (INIS)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)

  3. Commissioning and operation of DRAGON

    International Nuclear Information System (INIS)

    Engel, Sabine

    2003-01-01

    The new DRAGON (Detector of Recoils And Gammas Of Nuclear reactions) facility, located at the TRIUMF-ISAC radioactive beams laboratory in Vancouver, Canada, has initiated its experimental program. Recently DRAGON was used for initial studies of the 21 Na(p,γ) 22 Mg reaction. This facility was designed to measure absolutely the rates of radiative proton and alpha capture reactions of astrophysical interest to a precision of ±20%, using inverse kinematics. To fully understand the optics and operational parameters of the facility along with the transmission particularly of the reaction recoils, systematic studies of various configurations are in progress using stable beams along with measurements of well-known resonance reactions. The status of these commissioning studies is presented

  4. Commissioning and operation of DRAGON

    CERN Document Server

    Engel, S

    2003-01-01

    The new DRAGON (Detector of Recoils And Gammas Of Nuclear reactions) facility, located at the TRIUMF-ISAC radioactive beams laboratory in Vancouver, Canada, has initiated its experimental program. Recently DRAGON was used for initial studies of the sup 2 sup 1 Na(p,gamma) sup 2 sup 2 Mg reaction. This facility was designed to measure absolutely the rates of radiative proton and alpha capture reactions of astrophysical interest to a precision of +-20%, using inverse kinematics. To fully understand the optics and operational parameters of the facility along with the transmission particularly of the reaction recoils, systematic studies of various configurations are in progress using stable beams along with measurements of well-known resonance reactions. The status of these commissioning studies is presented.

  5. Graphite behaviour in relation to the fuel element design

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Manzel, R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Blackstone, R. [Reactor Centrum, Petten (Netherlands); Delle, W. [Kernforschungsanlage, Juelich (Germany); Lungagnani, V. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands); Krefeld, R. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands)

    1969-09-01

    The first designs of H.T.R. power reactors will probably use a Gilsocarbon based graphite for both the moderator/carrier blocks and for the fuel tubes. The initial physical properties and changes of dimensions, thermal expansion coefficient, Young*s modulus, and thermal conductivity on irradiation of Gilsocarbon graphites to typical reactor dwell-time fast neutron doses of 4 * 1021 cm -2 Ni dose Dido equivalent are given and values for the irradiation creep constant are presented. The influence of these property changes and those of chemical corrosion are considered briefly in relation to the present fuel element designs. The selection of an eventual less costly replacement graphite for Gilsocarbon graphite is discussed in terms of materials properties.

  6. Status of development of the HTR module

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1989-01-01

    Growing concern about the rising global temperature of the earth due to the ''Greenhouse Effect'' is increasingly focussing worldwide interest on passively safe reactors for heat and power production. In this context the development status of the HTR-Module designed by the Siemens-Group merits strong interest. The HTR-Module has a high degree of passive safety features. Even in case of hypothetical accidents the decay heat is dissipated from the primary system to the environment by passive measures alone i.e. by heat conduction, convection and radiation. The detailed engineering for the HTR-Module continues to progress. In addition to the engineering for the layout considerable progress has been made in the detailed engineering for specific components - e.g. pressure vessel, steam generator, hot gas duct, blower etc. - and specific systems - e.g. first core, helium purification system, reactor safety system, reactor control etc. The procedure for the conceptual licence has been continued. A large number of supplementary analyses and reports have been elaborated and submitted for this procedure. Many workshop meetings have been held with the nominated experts. The hypothetical accidents have been analysed and a special report on these accidents has been submitted. The safety analyses report has been revised, taking into account the results and achievements reached during the ongoing licensing procedure. Parallel to these engineering activities outstanding in R and D work for the HTR-Module, e.g. in the field of fuel elements etc. has been continued. The HTR-Module has found worldwide interest. Respective activities are going on in Bangladesh, PR China, USSR, Indonesia etc. Relevant application studies have been carried out and/or initiated. (author). 15 refs, 16 figs

  7. HTR-10GT AMBs displacement sensor design

    International Nuclear Information System (INIS)

    Shi Zhengang; Zha Meisheng; Zhao Lei; Sun Zhuo

    2005-01-01

    The 10 MW high temperature gas-cooled test module reactor (HTR-10GT) with the core made of spherical fuel elements was designed and constructed by the Institute of Nuclear and New Energy Technology of Tsinghua University in China. In the HTR-10GT, turbo-compressor and generator rotors are connected by a flexible coupling. The rotors, restricted by actual instruments and working environment, must be supported without any contact and lubrication. Active magnetic bearing (AMB), known as its advantages over the conventional bearings., such as contact-free, no-lubricating and active damping vibration, is the best way to suspend and stabilize the position of rotors of HTR-10GT. Each rotor is suspended by two radial and one axial AMBs. The radial AMB's radial gap is 0.15 mm considering the gap of 0.4 mm between the compressor stator and blades in order to protect the compressor. The control system controls the rotor position to meet the required gaps between rotor and stator through windings current. All the position information concerning radial and axial AMB is generated by sensors for measuring the displacement of the levitated body. Some typical sensors, i.e. eddy current displacement sensor, capacitive displacement sensor, can provide position information, but, quite often, unsatisfactory anti-jamming, which is a key issue for AMB systems near generator and other electric devices in HTR-10GT. Therefore, a kind of new type sensor is designed to measure the radial and axial displacements and the vibration of the rotors. This paper focuses on the design characteristics of the HTR-10GT AMBs displacement sensors and introduction of the related experiments to demonstrate its performance. (authors)

  8. An HTR cogeneration system for industrial applications

    International Nuclear Information System (INIS)

    Haverkate, B.R.W.; Heek, A.I. van; Kikstra, J.F.

    2001-01-01

    Because of its favourable characteristics of safety and simplicity the high-temperature reactor (HTR) could become a competitive heat source for a cogeneration unit. The Netherlands is a world leading country in the field of cogeneration. As nuclear energy remains an option for the medium and long term in this country, systems for nuclear cogeneration should be explored and developed. Hence, ECN Nuclear Research is developing a conceptual design of an HTR for Combined generation of Heat and Power (CHP) for the industry in and outside the Netherlands. The design of this small CHP-unit for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. The concept that was subject of this study, INCOGEN, used a 40 MW thermal pebble bed HTR and produced a maximum amount of electricity plus low temperature heat. The system has been improved to produce industrial quality heat, and has been renamed ACACIA. The output of this installation is 14 MW electricity and 17 tonnes of steam per hour, with a pressure of 10 bar and a temperature of 220 deg. C. The economic characteristics of this installation turned out to be much more favourable using modern data. The research work for this installation is embedded in a programme that has links to the major HTR projects in the world. Accordingly ECN participates in several IAEA Co-ordinated Research Programmes (CRPs). Besides this, ECN is involved in the South African PBMR-project. Finally, ECN participates in the European Concerted Action on Innovative HTR. (author)

  9. Means, methods and performances of the AREVA's HTR compact controls

    International Nuclear Information System (INIS)

    Banchet, J.; Guillermier, P.; Tisseur, D.; Vitali, M. P.

    2008-01-01

    In the AREVA's HTR development program, the reactor plant is composed of a prismatic core containing graphite cylindrical fuel elements, called compacts, where TRISO particles are dispersed. Starting from its past compacting process, the latter being revamped through the use of state of the art equipments, CERCA, 100% AREVA NP's subsidiary, was able to recover the quality of past compacts production. The recovered compacting process is composed of the following manufacturing steps: graphite matrix granulation, mix between the obtained granulates and particles, compacting and calcining at low pressure and temperature. To adapt this past process to new manufacturing equipments, non destructive examination tests were carried out to assess the compact quality, the latter being assessed via in house developed equipments and methods at each step of the design of experiments. As for the manufacturing process, past quality control methods were revamped to measure compact dimensional features (diameter, perpendicularity and cone effect), visual aspect, SiC layer failure fraction (via anodic disintegration and burn leach test) and homogeneity via 2D radiography coupled to ceramography. Although meeting quality requirements, 2D radiography method could not provide a quantified specification for compact homogeneity characterization. This limitation yielded the replacement of this past technique by a method based on X-Ray tomography. Development was conducted on this new technique to enable the definition of a criterion to quantify compact homogeneity, as well as to provide information about the distances in between particles. This study also included a comparison between simulated and real compacts to evaluate the accuracy of the technique as well as the influence of particle packing fraction on compact homogeneity. The developed quality control methods and equipments guided the choices of manufacturing parameters adjustments at the development stage and are now applied for

  10. Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

    International Nuclear Information System (INIS)

    Bess, John D.; Montierth, Leland; Köberl, Oliver

    2014-01-01

    Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235 U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  11. Advanced Control Techniques for WEC Wave Dragon

    DEFF Research Database (Denmark)

    Tedd, James; Kofoed, Jens Peter; Jasinski, M.

    2007-01-01

    This paper presents the ongoing work on control of the Wave Dragon wave energy converter. Research is being conducted in and between several centers across Europe. This is building upon the knowledge gained in the prototype project, and will enable much better performance of the future deployment...... of the full scale Wave Dragon....

  12. HTR-TN a European network for the development of HTR technology

    International Nuclear Information System (INIS)

    Von Lensa, W.

    2001-01-01

    A network called High-temperature reactor technology network (HTR-TN) has been created at a European level to coordinate works and knowledge on the subject with a long-term perspective and to serve as a channel for international collaboration. An analysis confirmed that the obvious economic penalty of HTR due to its low density power could be compensated by the combination of recent advances that may completely change the positioning of HTR on the energy market: -) the modular concept allowed to get a reactor free from core melt risk without intervention of any active safety system, implying a drastic simplification of the design of the reactor and the safety systems as well as a standardisation and potential for shop fabrication in series; -) the development of gas turbines, the efficiency of which increased, in 10 years, from 35% till 50% and more, enabling to consider suppression of the secondary system; -) the ultra high burn-up potential of HTR fuel and the possibility for direct disposal of spent HTR fuel elements that may reduce cost of the fuel cycle and contribute to the reduction of civil and military plutonium stockpiles. (A.C.)

  13. Outliers and Extremes: Dragon-Kings or Dragon-Fools?

    Science.gov (United States)

    Schertzer, D. J.; Tchiguirinskaia, I.; Lovejoy, S.

    2012-12-01

    Geophysics seems full of monsters like Victor Hugo's Court of Miracles and monstrous extremes have been statistically considered as outliers with respect to more normal events. However, a characteristic magnitude separating abnormal events from normal ones would be at odd with the generic scaling behaviour of nonlinear systems, contrary to "fat tailed" probability distributions and self-organized criticality. More precisely, it can be shown [1] how the apparent monsters could be mere manifestations of a singular measure mishandled as a regular measure. Monstrous fluctuations are the rule, not outliers and they are more frequent than usually thought up to the point that (theoretical) statistical moments can easily be infinite. The empirical estimates of the latter are erratic and diverge with sample size. The corresponding physics is that intense small scale events cannot be smoothed out by upscaling. However, based on a few examples, it has also been argued [2] that one should consider "genuine" outliers of fat tailed distributions so monstrous that they can be called "dragon-kings". We critically analyse these arguments, e.g. finite sample size and statistical estimates of the largest events, multifractal phase transition vs. more classical phase transition. We emphasize the fact that dragon-kings are not needed in order that the largest events become predictable. This is rather reminiscent of the Feast of Fools picturesquely described by Victor Hugo. [1] D. Schertzer, I. Tchiguirinskaia, S. Lovejoy et P. Hubert (2010): No monsters, no miracles: in nonlinear sciences hydrology is not an outlier! Hydrological Sciences Journal, 55 (6) 965 - 979. [2] D. Sornette (2009): Dragon-Kings, Black Swans and the Prediction of Crises. International Journal of Terraspace Science and Engineering 1(3), 1-17.

  14. Verification test of control rod system for HTR-10

    International Nuclear Information System (INIS)

    Zhou Huizhong; Diao Xingzhong; Huang Zhiyong; Cao Li; Yang Nianzu

    2002-01-01

    There are 10 sets of control rods and driving devices in 10 MW High Temperature Gas-cooled Test Reactor (HTR-10). The control rod system is the controlling and shutdown system of HTR-10, which is designed for reactor criticality, operation, and shutdown. In order to guarantee technical feasibility, a series of verification tests were performed, including room temperature test, thermal test, test after control rod system installed in HTR-10, and test of control rod system before HTR-10 first criticality. All the tests data showed that driving devices working well, control rods running smoothly up and down, random position settling well, and exactly position indicating

  15. Post-irradiation examination of a 13000C-HTR fuel experiment Project J 96.M3

    International Nuclear Information System (INIS)

    Bueger, J. de; Roettger, H.

    1977-01-01

    A large variety of loose coated fuel particles have been irradiated in the BR2 at Mol/Belgium at temperatures between 1200 0 C and 1400 0 C and up to a fast neutron fluence of 1.2x1022 cm -2 (E>0.1 MeV) as a Euratom sponsored experiment for the advanced testing of HTR fuel. The specimens have been provided by Belgonucleaire and the Dragon Project. A short description of the experiment as well as the results of post-irradiation examination mainly carried out at Petten (N.H.), The Netherlands, are presented here. The post-irradiation examination has shown that the required performance can be achieved by a number of the tested fuel specimens without serious damage

  16. Test facilities for HTR, (2)

    International Nuclear Information System (INIS)

    Ishizuka, Hiroshi; Hayakawa, Hitoshi; Miki, Toshiya.

    1981-01-01

    The core of the multi-purpose high temperature gas-cooled experimental reactor is a circular column as a whole, in which the columns of stacked graphite blocks of hexagonal prism are arranged. The blocks in a column are doweled so as not to move horizontally, but adjacent columns vibrate while colliding mutually at the time of an earthquake because there is a gap between them. For the purpose to know the vibrating characteristics of a column surrounded by gap, Fuji Electric Co., Ltd., carried out the experiment. The tested column, the testing setup and the test result are reported. The distribution of flow rate in the core must be clarified, and the design data must be established early for confirming the feasibility of core design. The core structure tester was installed in Japan Atomic Energy Research Institute. The 1/2.75 scale model of the reactor bed was used, and the sealing performance of the block assemblies was tested. The sealing tester is related also to the distribution of flow rate in the core, and the basic performance of seal elements and the cross flow in fuel blocks were tested. The one-column tester and the seal element/two-column tester, the piping unit and the blower filter unit compose this tester. (Kako, I.)

  17. PCTR experiments with HTR lattice in MARIUS

    Energy Technology Data Exchange (ETDEWEB)

    Gambier, G; Estiot, J C; de Lapperent, D; Laponche, B; Luffin, J; Morier, F

    1972-06-15

    PCTR experiments have been carried out in Marius III with HTR tubular fuel, enriched to around 1% in order to reach K{sub infinity} = 1 and to reduce the mass of poison. Three poisons were used - Aluminium, Copper and Vanadium. The effect of air was measured and corrections were made to the results to allow the effect of delayed neutrons and the effect of axial heterogeneities. Interpretation was made with APOLLO. (auth)

  18. Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

    International Nuclear Information System (INIS)

    Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

    1977-03-01

    Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

  19. The HTR, applications, economics and environmental aspects

    International Nuclear Information System (INIS)

    Barnert, H.; Schad, M.; Candeli, H.

    1990-01-01

    The High Temperature Reactor (HTR), as the only nuclear system producing high temperature heat up to 1000 deg. C, offers a wide variety of applications. Besides electricity production, via steam turbines and in future via gas turbines, there is: District heat with high efficiency, long distance energy for urban energy supply, high pressure injection steam production for enhanced oil recovery, medium range temperature heat direct application in chemical and related industry and last not least, high temperature application for the refinement of fossil energy carriers. Recent results of studies and programmes will be presented: Near term applications are identified, e.g. refineries and alumina industry with smaller HTR units. Another large market is the production of hydrogen, methanol and ammonia on the basis of natural gas, the relevant technology has been developed up to the pilot scale. The refinement of fossil energy carriers, in particular of coal, is subject of the R+D programme in the cooperation between German industrial companies and the Nuclear Research Center. The results are very promising and will be explained in detail. This programme will be continued. Objectives are: improvement of the technology and of the economics as well as environmental aspects, e.g. the reduction of emissions of carbon-dioxid. The topics of the programme deal with the different apparatus, e.g. steam methane reformer, steam coal gasifier, intermediate heat exchanger and last not least, the process heat HTR. (author)

  20. An HTR cogeneration system for industrial application

    International Nuclear Information System (INIS)

    Haverkate, B.R.W.; Van Heek, A.I.; Kikstra, J.F.

    1999-01-01

    Because of its favourable characteristics of safety and simplicity the high-temperature reactor (HTR) could become a competitive heat source for a cogeneration unit. The Netherlands is a world leading country in the field of cogeneration. As nuclear energy remains an option for the medium and long term in this country, systems for nuclear cogeneration should be explored and developed. Hence, ECN Nuclear Research is developing a conceptual design of an HTR for Combined generation of Heat and Power (CHP) for the industry in and outside the Netherlands. The design of this small CHP-unit for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. The concept that was subject of that study, INCOGEN, used a 40 MW thermal pebble bed HTR and produced a maximum amount of electricity plus low temperature heat. The system has been improved to produce industrial quality heat, and has been renamed ACACIA. The output of this installation is 14 MW electricity and 17 tonnes of steam per hour, with a pressure of 10 bar and a temperature of 220C. The economic characteristics of this installation turned out to be much more favourable using modern cost data. 15 refs

  1. Potentialities of high temperature reactors (HTR)

    International Nuclear Information System (INIS)

    Hittner, D.

    2001-01-01

    This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)

  2. Experiments with the Dragon Machine

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    2005-01-01

    The basic characteristics of a self-sustaining chain reaction were demonstrated with the Chicago Pile in 1943, but it was not until early 1945 that sufficient enriched material became available to experimentally verify fast-neutron cross-sections and the kinetic characteristics of a nuclear chain reaction sustained with prompt neutrons alone. However, the demands of wartime and the rapid decline in effort following the cessation of hostilities often resulted in the failure to fully document the experiments or in the loss of documentation as personnel returned to civilian pursuits. When documented, the results were often highly classified. Even when eventually declassified, the data were often not approved for public release until years later.2 Even after declassification and approval for public release, the records are sometimes difficult to find. Through a fortuitous discovery, a set of handwritten notes by ''ORF July 1945'' entitled ''Dragon - Research with a Pulsed Fission Reactor'' was found by William L. Myers in an old storage safe at Pajarito Site of the Los Alamos National Laboratory3. Of course, ORF was identified as Otto R. Frisch. The document was attached to a page in a nondescript spiral bound notebook labeled ''494 Book'' that bore the signatures of Louis Slotin and P. Morrison. The notes also reference an ''Idea LS'' that can only be Louis Slotin. The discovery of the notes led to a search of Laboratory Archives, the negative files of the photo lab, and the Report Library for additional details of the experiments with the Dragon machine that were conducted between January and July 1945. The assembly machine and the experiments were carefully conceived and skillfully executed. The analyses--without the crutch of computers--display real insight into the characteristics of the nuclear chain reaction. The information presented here provides what is believed to be a complete collection of the original documentation of the observations made with the Dragon

  3. A graphite foam reinforced by graphite particles

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

    2007-11-15

    Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

  4. Operational experience with Dragon reactor experiment of relevance to commercial reactors

    International Nuclear Information System (INIS)

    Capp, P.D.; Simon, R.A.

    1976-01-01

    An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

  5. HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  6. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  7. A synthesis on the HTR scenario studies at CEA - HTR2008-58059

    International Nuclear Information System (INIS)

    Boucher, L.; Greneche, D.

    2008-01-01

    The aim of the studies is to assess the impact of the deployment of an HTR park replacing one part of the current PWR reactors. The other part of the current park is replaced by EPRs. In these scenarios, the annual electricity production is constant at 400 TWhe. This value corresponds roughly to the present nuclear electricity production in France. From 2002 to 2007, an important program study on HTR has been carried out by CEA and AREVA NC under the joint CEA - AREVA NC project 'prospective studies on the management of Plutonium and the back end of the cycle'. This program addresses core physic and scenario studies, and also the back end of the fuel cycle : reprocessing of spent fuel and HTR waste management. Some core physic studies have already been presented in the reference [1]. This paper presents the results of the scenario studies using two concepts: either the standard core of the Gas Turbine Modular Helium Reactor concept (GTMHR) with Uranium or Plutonium fuel, or the Multiple Fuel Rows Core (MFRC) dedicated to the actinide burning. The insertion of a new concept (fuel, reactor, process) must be evaluated in the global electronuclear system with an analysis of the impact on the fuel cycle (Enrichment, Fuel Fabrication, Reactor, Processing, Interim Storage, Waste storage). The scenario studies are used to evaluate different solutions to manage nuclear materials (uranium, plutonium) and wastes (minor actinides and fission products), from the present situation in France (closed cycle with storage of used MOX fuels) until the final equilibrium: mixed nuclear park with EPR and HTR. These studies allow to calculate material flows and inventories of these elements in each step of the fuel cycle. The simulation of transient scenarios from the present situation to the future situation is performed with the COSI code. HTR reactors feature a high flexibility with regard to fuel cycle options. Several versions of core have been investigated, with different type of

  8. The Energy Conversion Analysis of HTR Gas Turbine System

    International Nuclear Information System (INIS)

    Utaja

    2000-01-01

    The energy conversion analysis of HTR gas turbine system by hand calculation is tedious work and need much time. This difficulty comes from the repeated thermodynamic process calculation, both on compression or expansion of the cycle. To make the analysis faster and wider variable analyzed, HTR-1 programme is used. In this paper, the energy conversion analysis of HTR gas turbine system by HTR-1 will be described. The result is displayed as efficiency curve and block diagram with the input and output temperature of the component. This HTR-1 programme is developed by Basic language programming and be compiled by Visual Basic 5.0 . By this HTR-1 programme, the efficiency, specific power and effective compression of the amount of gas can be recognized fast. For example, for CO 2 gas between 40 o C and 700 o C, the compression on maximum efficiency is 4.6 and the energy specific is 18.9 kcal/kg, while the temperature changing on input and output of the component can be traced on monitor. This process take less than one second, while the manual calculation take more than one hour. It can be concluded, that the energy conversion analysis of the HTR gas turbine system by HTR-1 can be done faster and more variable analyzed. (author)

  9. Transport of fission products in matrix and graphite

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1983-06-01

    In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

  10. PEMODELAN TERAS UNTUK ANALISIS PERHITUNGAN KONSTANTA MULTIPLIKASI REAKTOR HTR-PROTEUS

    Directory of Open Access Journals (Sweden)

    Zuhair Zuhair

    2015-04-01

    with packing fraction of 5.76% without exclusive zone. Pebble bed core modeling was approximated by utilizing regular lattice of balls that are arranged as BCC lattice based on repeated cell generated from a numerous unit cell. The MCNP5 calculation results showed that excellent agreement with the experiment, although the HTRPROTEUS core predicted more reactive than the measurement, especially in cores 4.2 and 4.3. ENDF/B-VI library indicates consistency with the most accurate keff estimation compared to ENDF/B-V library, mainly ENDF/B-VI (66c. The deviation of calculated keff estimation with experiment is attributed to the consequence of specified graphite reflector composition. The comparison conducted shows that MCNP5 produces HTR-PROTEUS core keff is more precise compared to the results of MCNP4B and MCNP-BALL. These results concluded that the success of this modeling methodology justifies MCNP5 application for other pebble bed reactor analysis. Keywords: HTR-PROTEUS core modeling multiplication constant, MCNP5

  11. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2011-10-01

    MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  12. Tardive dyskinesia and DRD3, HTR2A and HTR2C gene polymorphisms in Russian psychiatric inpatients from Siberia

    NARCIS (Netherlands)

    Al Hadithy, A. F. Y.; Ivanova, S. A.; Pechlivanoglou, P.; Semke, A.; Fedorenko, O.; Kornetova, E.; Ryadovaya, L.; Brouwers, J. R. B. J.; Wilffert, B.; Bruggeman, R.; Loonen, A. J. M.

    2009-01-01

    Background: Pharmacogenetics of tardive dyskinesia and dopamine D3 (DRD3), serotonin 2A (HTR2A), and 2C (HTR2C) receptors has been examined in various populations, but not in Russians. Purpose: To investigate the association between orofaciolingual (TDof) and limb-truncal dyskinesias (TDlt) and

  13. HTR-PM Safety requirement and Licensing experience

    International Nuclear Information System (INIS)

    Li Fu; Zhang Zuoyi; Dong Yujie; Wu Zongxin; Sun Yuliang

    2014-01-01

    HTR-PM is a 200MWe modular pebble bed high temperature reactor demonstration plant which is being built in Shidao Bay, Weihai, Shandong, China. The main design parameters of HTR-PM were fixed in 2006, the basic design was completed in 2008. The review of Preliminary Safety Analysis Report (PSAR) of HTR-PM was started in April 2008, completed in September 2009. In general, HTR- PM design complies with the current safety requirement for nuclear power plant in China, no special standards are developed for modular HTR. Anyway, Chinese Nuclear Safety Authority, together with the designers, developed some dedicated design criteria for key systems and components and published the guideline for the review of safety analysis report of HTR-PM, based on the experiences from licensing of HTR-10 and new development of nuclear safety. The probabilistic safety goal for HTR-PM was also defined by the safety authority. The review of HTR-PM PSAR lasted for one and a half years, with 3 dialogues meetings and 8 topics meetings, with more than 2000 worksheets and answer sheets. The heavily discussed topics during the PSAR review process included: the requirement for the sub-atmospheric ventilation system, the utilization of PSA in design process, the scope of beyond design basis accidents, the requirement for the qualification of TRISO coating particle fuel, and etc. Because of the characteristics of first of a kind for the demonstration plant, the safety authority emphasized the requirement for the experiment and validation, the PSAR was licensed with certain licensing conditions. The whole licensing process was under control, and was re-evaluated again after Fukushima accident to be shown that the design of HTR-PM complies with current safety requirement. This is a good example for how to license a new reactor. (author)

  14. Dungeons & Dragons: The gamers are revolting! [symposium

    Directory of Open Access Journals (Sweden)

    Rebecca Bryant

    2009-03-01

    Full Text Available The negative response by players to corporate changes to the rule systems governing Dungeons & Dragons suggests that tabletop RPGs have more in common with fan fiction than with computer games.

  15. Wave Overtopping Characteristics of the Wave Dragon

    DEFF Research Database (Denmark)

    Tedd, James; Kofoed, Jens Peter

    Simulation work has been used extensively with the Wave dragon and other overtopping devices to analyse the power production performance of them and to optimise the structural design and the control strategy. A time domain approach to this is well documented in Jakobsen & Frigaard 1999. Using...... measurements taken from the Wave Dragon Nissum Bredning prototype, some of the previous assumptions have been slightly modified and improved upon, so that the simulation method better represents the reality of what is occurring....

  16. Thermally exfoliated graphite oxide

    Science.gov (United States)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Abdala, Ahmed (Inventor)

    2011-01-01

    A modified graphite oxide material contains a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the thermally exfoliated graphite oxide displays no signature of the original graphite and/or graphite oxide, as determined by X-ray diffraction.

  17. DEM simulation of particle mixing for optimizing the overcoating drum in HTR fuel fabrication

    Science.gov (United States)

    Liu, Malin; Lu, Zhengming; Liu, Bing; Shao, Youlin

    2013-06-01

    The rotating drum was used for overcoating coated fuel particles in HTR fuel fabrication process. All the coated particles should be adhered to equal amount of graphite powder, which means that the particle should be mixed quickly in both radial and axial directions. This paper investigated the particle flow dynamics and mixing behavior in different regimes using the discrete element method (DEM). By varying the rotation speed, different flow regimes such as slumping, rolling, cascading, cataracting, centrifuging were produced. The mixing entropy based on radial and axial grid was introduced to describe the radial and axial mixing behaviors. From simulation results, it was found that the radial mixing can be achieved in the cascading regime more quickly than the slumping, rolling and centrifuging regimes, but the traditional rotating drum without internal components can not achieve the requirements of axial mixing and should be improved. Three different structures of internal components are proposed and simulated. The new V-shaped deflectors were found to achieve a quick axial mixing behavior and uniform axial distribution in the rotating drum based on simulation results. At last, the superiority was validated by experimental results, and the new V-shaped deflectors were used in the industrial production of the overcoating coated fuel particles in HTR fuel fabrication process.

  18. The improvement of the method of equivalent cross section in HTR

    International Nuclear Information System (INIS)

    Guo, J.; Li, F.

    2012-01-01

    The Method of Equivalence Cross-Sections (MECS) is a combined transport-diffusion method. By appropriately adjusting the diffusion coefficient of homogenized absorber region, the diffusion theory could yield satisfactory results for the full core model with strong neutron absorber material, for example the control rod in High temperature gas cooled reactor (HTR). Original implementation of MECS based on 1-D cell transport model has some limitation on accuracy and applicability, a new implementation of MECS based on 2-D transport model are proposed and tested in this paper. This improvement can extend the MECS to the calculation of twin small absorber ball system which have a non-circular boring in graphite reflector and different radial position. A least-square algorithm for the calculation of equivalent diffusion coefficient is adopted, and special treatment for diffusion coefficient for higher energy group is proposed in the case that absorber is absent. Numerical results to adopt MECS into control rod calculation in HTR are encouraging. However, there are some problems left. (authors)

  19. Effects of homogeneous geometry models in simulating the fuel balls in HTR-10

    International Nuclear Information System (INIS)

    Wang Mengjen; Liang Jenqhorng; Peir Jinnjer; Chao Dersheng

    2012-01-01

    In this study, the core geometry of HTR-10 was simulated using four different models including: (1) model 1 - an explicit double heterogeneous geometry, (2) model 2 - a mixing of UO 2 kernel and four layers in each TRISO particle into one, (3) model 3 - a mixing of 8,335 TRISO particles and the inner graphite matrix in each fuel ball into one, and (4) model 4 - a mixing of the outer graphite shell, 8,335 TRISO particles, and the inner graphite matrix in each fuel ball into one. The associated initial core computations were performed using the MCNP version 1.51 computer code. The experimental fuel loading height of 123 cm was employed for each model. The results revealed that the multiplication factors ranged from largest to smallest with model 1, model 2, model 3, and model 4. The neutron spectrum in the fuel region of each models varied from the hardest to the softest are model 1, model 2, model 3, and model 4 while the averaged neutron spectrum in fuel ball from hardest to softest are model 4, model 3, model 2, and model 1. In addition, the CPU execution times extended from longest to shortest with model 1, model 2, model 3, and model 4. (author)

  20. Plutonium re-cycle in HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J. A.

    1974-03-15

    The study of plutonium cycles in HTRs using reprocessed plutonium from Magnox and AGR fuel cycles has shown that full core plutonium/uranium loadings are in general not feasible, burn-up is limited due the need for lower loadings of plutonium to meet reload core reactivity limits, on-line refueling is not practicable due to the need for higher burnable poison loadings, and low conversion rates in the plutonium-uranium cycles cannot be mitigated by axial loading schemes so that fissile make-up is needed if HTR plutonium recycle is desired.

  1. Dragon bridge - the world largest dragon-shaped (ARCH steel bridge as element of smart city

    Directory of Open Access Journals (Sweden)

    Chinh Luong Minh

    2016-01-01

    Full Text Available Dragon Bridge - The world’s largest dragon-shaped steel bridge, with an installation cost of $85 million USD, features 6 lanes for two separate directions, 666 meters of undulating steel in the shape of a dragon in the Ly Dynasty, the symbol of prosperity in Vietnamese culture. This unique and beautifully lit bridge, which also breathes fire and sprays water. It’s the purposeful integration of the lighting hardware articulates the dragon’s form, and the fire-breathing dragon head. This project transcends the notion of monumental bridge with dynamic colour-changing lighting, creating an iconic sculpture in the skyline that is both reverent and whimsical. The signature feature of the bridge was the massive undulating support structure resembling a dragon flying over the river. The dragon is prominent in Vietnamese culture as a symbol of power and nobility. Dragon Bridge stands out as a model of innovation. It has received worldwide attention in the design community and from the global media for its unique arch support system. Dragon Bridge serves as an example of how aesthetic quality of a design can serve cultural, economic and functional purposes. The article presents design solutions of the object and the evaluation of the technical condition before putting the facility into service.

  2. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  3. The use and development of the high-temperature reactor (HTR) in China. A conference report

    International Nuclear Information System (INIS)

    Marnet, C.

    2001-01-01

    Gas-cooled graphite-moderated reactors have been under development since the early days of nuclear technology. Starting with plants in Britain and France, reactors employing this combination of coolant and moderator were used in commercial nuclear power plants in the second half of the fifties. At the same time, efforts seeking to use inert helium gas as a coolant resulted in the construction in several countries, the United States and Germany in particular, of larger nuclear power plants with higher coolant temperatures and the resultant thermodynamic advantages of high efficiencies and the option of process heat generation. Economic and political considerations led to the decommissioning of these plants. Today, research and development of high-temperature reactors are concentrated on smaller units. Work is carried out in close international cooperation, especially on project designs and newly commissioned plants in China (HTR-10), Japan (HTTR, Oarai), And South Africa (ESKOM project), but also in the USA and in Russia. (orig.) [de

  4. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach; Simulacao termohidraulica do nucleo do reator nuclear HTR-10 com o uso da abordagem realistica CFD

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S.; Dominguez, Dany S., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil); Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba); Lira, Carlos Alberto Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  5. Bridged graphite oxide materials

    Science.gov (United States)

    Herrera-Alonso, Margarita (Inventor); McAllister, Michael J. (Inventor); Aksay, Ilhan A. (Inventor); Prud'homme, Robert K. (Inventor)

    2010-01-01

    Bridged graphite oxide material comprising graphite sheets bridged by at least one diamine bridging group. The bridged graphite oxide material may be incorporated in polymer composites or used in adsorption media.

  6. The HTR-10 project and its further development

    International Nuclear Information System (INIS)

    Xu Yuanhui

    2002-01-01

    The 10 MW High Temperature Gas-cooled Reactor-Test Module (termed as HTR-10) is one of key project in the National High Technology Research and Development Program (1986-2000). Main objectives for the HTR-10 are: (1). To acquire know-how to design, construct and operate the HTGRs, (2). To establish an experimental facility, (3). To demonstrate the inherent safety features of the Modular HTGR, (4). To test electricity and heat co-generation and closed cycle gas turbine technology and (5). To do research and development work for high temperature process heat application. The Institute of Nuclear Energy Technology (INET) of Tsinghua University was appointed as the leading institute to be responsible for design, license applications, construction and operation of the HTR-10. The HTR-10 technical design represents the features of HTR-Module design. After five years construction, installation and pre-operation the HTR-10 reached the criticality in December 2000. Up to now all of results on zero point experiments and fuel elements irradiation test are fine. China will continue to develop the high temperature gas-cooled reactor in the future using the HTR-10 base

  7. Evaluation of the Control Rod Super Alloy Material of HTR-PM

    International Nuclear Information System (INIS)

    Li Pengjun; Yan He; Diao Xingzhong

    2014-01-01

    The control rod drive mechanism (CRDM) system is served as the first reactivity control and shutdown system for the high temperature reactor pebble-bed module (HTR-PM) in Shandong, China. And the control rod, which is pulled up and down by a chain sprocket mechanism of CRDM to realize reactivity control, compensation and shutdown, has to be durable under temperature as high as 550℃ for a long time. Thus the material persistent strength under high temperature is quite important for the reliability of the CRDM. In this paper, a review on material selection of control rod of high temperature gas cooled reactors, including AVR and THTR-300 in Germany, HTTR in Japan, PBMR in South Africa and Dragon in Britain, was summarized. The major parameters of two kinds of high temperature alloy, incoloy 800H and alloy 625, were compared and discussed. According to the ASME NH volume, a design criterion for the control rod was established and applied in the analysis of the chain by using finite element method. The numerical simulations showed that the chain made of alloy 625 could meet the condition and work for a long time under high temperature. (author)

  8. Process for purifying graphite

    International Nuclear Information System (INIS)

    Clausius, R.A.

    1985-01-01

    A process for purifying graphite comprising: comminuting graphite containing mineral matter to liberate at least a portion of the graphite particles from the mineral matter; mixing the comminuted graphite particles containing mineral matter with water and hydrocarbon oil to form a fluid slurry; separating a water phase containing mineral matter and a hydrocarbon oil phase containing grahite particles; and separating the graphite particles from the hydrocarbon oil to obtain graphite particles reduced in mineral matter. Depending upon the purity of the graphite desired, steps of the process can be repeated one or more times to provide a progressively purer graphite

  9. Structural and Functional Analysis of Human HtrA3 Protease and Its Subdomains.

    Directory of Open Access Journals (Sweden)

    Przemyslaw Glaza

    Full Text Available Human HtrA3 protease, which induces mitochondria-mediated apoptosis, can be a tumor suppressor and a potential therapeutic target in the treatment of cancer. However, there is little information about its structure and biochemical properties. HtrA3 is composed of an N-terminal domain not required for proteolytic activity, a central serine protease domain and a C-terminal PDZ domain. HtrA3S, its short natural isoform, lacks the PDZ domain which is substituted by a stretch of 7 C-terminal amino acid residues, unique for this isoform. This paper presents the crystal structure of the HtrA3 protease domain together with the PDZ domain (ΔN-HtrA3, showing that the protein forms a trimer whose protease domains are similar to those of human HtrA1 and HtrA2. The ΔN-HtrA3 PDZ domains are placed in a position intermediate between that in the flat saucer-like HtrA1 SAXS structure and the compact pyramidal HtrA2 X-ray structure. The PDZ domain interacts closely with the LB loop of the protease domain in a way not found in other human HtrAs. ΔN-HtrA3 with the PDZ removed (ΔN-HtrA3-ΔPDZ and an N-terminally truncated HtrA3S (ΔN-HtrA3S were fully active at a wide range of temperatures and their substrate affinity was not impaired. This indicates that the PDZ domain is dispensable for HtrA3 activity. As determined by size exclusion chromatography, ΔN-HtrA3 formed stable trimers while both ΔN-HtrA3-ΔPDZ and ΔN-HtrA3S were monomeric. This suggests that the presence of the PDZ domain, unlike in HtrA1 and HtrA2, influences HtrA3 trimer formation. The unique C-terminal sequence of ΔN-HtrA3S appeared to have little effect on activity and oligomerization. Additionally, we examined the cleavage specificity of ΔN-HtrA3. Results reported in this paper provide new insights into the structure and function of ΔN-HtrA3, which seems to have a unique combination of features among human HtrA proteases.

  10. Structural and Functional Analysis of Human HtrA3 Protease and Its Subdomains.

    Science.gov (United States)

    Glaza, Przemyslaw; Osipiuk, Jerzy; Wenta, Tomasz; Zurawa-Janicka, Dorota; Jarzab, Miroslaw; Lesner, Adam; Banecki, Bogdan; Skorko-Glonek, Joanna; Joachimiak, Andrzej; Lipinska, Barbara

    2015-01-01

    Human HtrA3 protease, which induces mitochondria-mediated apoptosis, can be a tumor suppressor and a potential therapeutic target in the treatment of cancer. However, there is little information about its structure and biochemical properties. HtrA3 is composed of an N-terminal domain not required for proteolytic activity, a central serine protease domain and a C-terminal PDZ domain. HtrA3S, its short natural isoform, lacks the PDZ domain which is substituted by a stretch of 7 C-terminal amino acid residues, unique for this isoform. This paper presents the crystal structure of the HtrA3 protease domain together with the PDZ domain (ΔN-HtrA3), showing that the protein forms a trimer whose protease domains are similar to those of human HtrA1 and HtrA2. The ΔN-HtrA3 PDZ domains are placed in a position intermediate between that in the flat saucer-like HtrA1 SAXS structure and the compact pyramidal HtrA2 X-ray structure. The PDZ domain interacts closely with the LB loop of the protease domain in a way not found in other human HtrAs. ΔN-HtrA3 with the PDZ removed (ΔN-HtrA3-ΔPDZ) and an N-terminally truncated HtrA3S (ΔN-HtrA3S) were fully active at a wide range of temperatures and their substrate affinity was not impaired. This indicates that the PDZ domain is dispensable for HtrA3 activity. As determined by size exclusion chromatography, ΔN-HtrA3 formed stable trimers while both ΔN-HtrA3-ΔPDZ and ΔN-HtrA3S were monomeric. This suggests that the presence of the PDZ domain, unlike in HtrA1 and HtrA2, influences HtrA3 trimer formation. The unique C-terminal sequence of ΔN-HtrA3S appeared to have little effect on activity and oligomerization. Additionally, we examined the cleavage specificity of ΔN-HtrA3. Results reported in this paper provide new insights into the structure and function of ΔN-HtrA3, which seems to have a unique combination of features among human HtrA proteases.

  11. Characterization of dragon fruit (Hylocereus spp.) components with valorization potential

    OpenAIRE

    Liaotrakoon, Wijitra

    2013-01-01

    Dragon fruit (Hylocereus spp.), also known as pitaya or pitahaya, is increasingly gaining interest in many countries, including Thailand which is a country with a climate ideal for breeding different varieties of tropical and subtropical fruits in general, and dragon fruit more specifically. The benefits of dragon fruit for human health can be explained by its essential nutrients such as vitamins, minerals, complex carbohydrates, dietary fibres and antioxidants. Dragon fruit is also an essent...

  12. The retardation effect of structural graphite on the release of fission products in case of hypothetical accidents of HTRs

    International Nuclear Information System (INIS)

    Iniotakis, N.; Decken, C.B. von der

    1982-01-01

    In case of a hypothetical core heat up accident of an HTR the structural graphite of the reactor causes under certain circumstances a very important retardation of the release of fission products into the containment building of the plant. A model is presented which describes the transport phenomena in the graphite structure extensively taking into account specially the macro-structure of the graphite. It is shown by parameter variations under which conditions one can expect a large retardation effect and quantitative values of this retardation, which can be very important, are given. (author)

  13. Instrumentation of steam cycle HTR's up to 900 MWe

    International Nuclear Information System (INIS)

    Leithner, D.E.; Winkenbach, B.

    1982-06-01

    Due to basic design features and inherent safety qualities in-core instrumentation is not needed in an HTR. Reactor safety requirements can be met by integral measurements. A modest spatial resolving power of the out-of-core instrumentation is sufficient for all operational purposes in small and medium sized steam cycle HTR's. Thus, the instrumentation concept of the THTR 300 MWe prototype reactor can be adopted without major changes for the HTR 450 MWe reactor project, as is demonstrated here for the neutron flux and temperature measurements. (author)

  14. A 350 MW HTR with an annular pebble bed core

    International Nuclear Information System (INIS)

    Wang Dazhong; Jiang Zhiqiang; Gao Zuying; Xu Yuanhui

    1992-12-01

    A conceptual design of HTR-module with an annular pebble bed core was proposed. This design can increase the unit power capacity of HTR-Module from 200 MWt to 350 MWt while it can keep the inherent safety characteristics of modular reactor. The preliminary safety analysis results for 350 MW HTR are given. In order to solve the problem of uneven helium outlet temperature distribution a gas flow mixing structure at bottom of core was designed. The experiment results of a gas mixing simulation test rig show that the mixing function can satisfy the design requirements

  15. AREVA HTR concept for near-term deployment

    Energy Technology Data Exchange (ETDEWEB)

    Lommers, L.J., E-mail: lewis.lommers@areva.com [AREVA Inc., 2101 Horn Rapids Road, Richland, WA 99354 (United States); Shahrokhi, F. [AREVA Inc., Lynchburg, VA (United States); Mayer, J.A. [AREVA Inc., Marlborough, MA (United States); Southworth, F.H. [AREVA Inc., Lynchburg, VA (United States)

    2012-10-15

    This paper introduces AREVA's High Temperature Reactor (HTR) steam cycle concept for near-term industrial deployment. Today, nuclear power primarily impacts only electricity generation. The process heat and transportation fuel sectors are completely dependent on fossil fuels. In order to impact this energy sector as rapidly as possible, AREVA has focused its HTR development effort on the steam cycle HTR concept. This reduces near-term development risk and minimizes the delay before a useful contribution to this sector of the energy economy can be realized. It also provides a stepping stone to longer term very high temperature concepts which might serve additional markets. A general description of the current AREVA steam cycle HTR concept is provided. This concept provides a flexible system capable of serving a variety of process heat and cogeneration markets in the near-term.

  16. Two Phase Flow Stability in the HTR-10 Steam Generator

    Institute of Scientific and Technical Information of China (English)

    居怀明; 左开芬; 刘志勇; 徐元辉

    2001-01-01

    A 10 MW High Temperature Gas Cooled Reactor (HTR-10) designed bythe Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.

  17. HTR plus modern turbine technology for higher efficiencies

    International Nuclear Information System (INIS)

    Barnert, H.; Kugeler, K.

    1996-01-01

    The recent efficiency race for natural gas fired power plants with gas-plus steam-turbine-cycle, is shortly reviewed. The question 'can the HTR compete with high efficiencies?' is answered: Yes, it can - in principle. The gas-plus steam-turbine cycle, also called combi-cycle, is proposed to be taken into consideration here. A comparative study on the efficiency potential is made; it yields 54.5% at 1,050 deg. C gas turbine-inlet temperature. The mechanisms of release versus temperature in the HTR are summarized from the safety report of the HTR MODUL. A short reference is made to the experiences from the HTR-Helium Turbine Project HHT, which was performed in the Federal Republic of Germany in 1968 to 1981. (author). 8 figs,. 1 tab

  18. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Klippel, H.T.; Hogenbirk, A.; Oppe, J.; Sciolla, C.M.; Stad, R.C.L. van der; Zhang, B.C.

    1997-06-01

    As part of the activities within the framework of the development of INCOGEN, a 'Dutch' conceptual design of a smaller HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRs, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (orig.)

  19. AREVA HTR concept for near-term deployment

    International Nuclear Information System (INIS)

    Lommers, L.J.; Shahrokhi, F.; Mayer, J.A.; Southworth, F.H.

    2012-01-01

    This paper introduces AREVA's High Temperature Reactor (HTR) steam cycle concept for near-term industrial deployment. Today, nuclear power primarily impacts only electricity generation. The process heat and transportation fuel sectors are completely dependent on fossil fuels. In order to impact this energy sector as rapidly as possible, AREVA has focused its HTR development effort on the steam cycle HTR concept. This reduces near-term development risk and minimizes the delay before a useful contribution to this sector of the energy economy can be realized. It also provides a stepping stone to longer term very high temperature concepts which might serve additional markets. A general description of the current AREVA steam cycle HTR concept is provided. This concept provides a flexible system capable of serving a variety of process heat and cogeneration markets in the near-term.

  20. HTR plus modern turbine technology for higher efficiencies

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-08-01

    The recent efficiency race for natural gas fired power plants with gas-plus steam-turbine-cycle, is shortly reviewed. The question `can the HTR compete with high efficiencies?` is answered: Yes, it can - in principle. The gas-plus steam-turbine cycle, also called combi-cycle, is proposed to be taken into consideration here. A comparative study on the efficiency potential is made; it yields 54.5% at 1,050 deg. C gas turbine-inlet temperature. The mechanisms of release versus temperature in the HTR are summarized from the safety report of the HTR MODUL. A short reference is made to the experiences from the HTR-Helium Turbine Project HHT, which was performed in the Federal Republic of Germany in 1968 to 1981. (author). 8 figs,. 1 tab.

  1. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Hass, J.B.M. De; Klippel, H.Th.; Hogenbirk, A.; Oppe, J.; Sciolla, C.; Stad, R.C.L. Van Der; Zhang, B.C.

    1997-01-01

    As part of the activities within the framework of the development of INCOGEN, a ''Dutch'' conceptual design of a small HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRS, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (author)

  2. How not to train your dragon: a case of a Komodo dragon bite.

    Science.gov (United States)

    Borek, Heather A; Charlton, Nathan P

    2015-06-01

    Komodo dragons (Varanus komodoensis) are the world's largest lizards, known for killing prey that exceed their body mass. Reports of bites to humans in the popular press suggest high degrees of morbidity and mortality. Reports in the medical literature are lacking. We describe the case of a zookeeper who was bitten by a Komodo dragon, with a resultant mallet finger. We further discuss the various potential mechanisms of Komodo dragon lethality, including sepsis and venom deposition theories that are useful in guiding management. Copyright © 2015 Wilderness Medical Society. Published by Elsevier Inc. All rights reserved.

  3. HTR-PM Progress and Further Commercial Deployment

    International Nuclear Information System (INIS)

    Wu, Frank

    2017-01-01

    Project Milestones: • 2004: industry investment agreement was signed • 2006: decided to use 2×250 MWt reactor modules with a 200 MWe steam turbine, became a key government R&D project • 2008: ATP was issued • 2012.12.9: FCD the first concrete poured. Chinese HTR development: HTR Roles in China - Power generation: supplement to LWR; repowering coal fired plants - Co-generation to supply steam - Hydrogen production

  4. Market potential of heat utilization of modular HTR in Japan

    International Nuclear Information System (INIS)

    Ide, Akira; Tasaka, Kanji.

    1993-01-01

    HTR is considered to be the most suitable reactor type to use in the field other than power generation. So it is useful to know market potential of this type of reactor in Japan to justify its development. This potential was estimated to be about 400 200MWt modular HTR reactors. This number will be double if the market of hydrogen is developed. (J.P.N.)

  5. Dragon-I Linear Induction Electron Accelerator

    International Nuclear Information System (INIS)

    Ding Bonan; Deng Jianjun; Wang Huacen; Cheng Nian'an; Dai Guangsen; Zhang Linwen; Liu Chengjun; Zhang Wenwei; Li Jin; Zhang Kaizhi

    2005-01-01

    Dragon-I is a linear induction electron accelerator. This facility consists of a 3.6 MeV injector, 38 meter beam transport line and 16 MeV induction accelerator powered by high voltage generators, including 8 Marx generators and 48 Blumlein lines. This paper describes the physics design, development and experimental results of Dragon-I. The key technology is analyzed in the accelerator development, and the design requirements and operation of the major subsystems are presented. The experimental results show Dragon-I generates an 18-20 MeV, 2.5 kA, 70 ns electron beam. The X-ray spot size is about 1.2 mm and dose level about 0.103 C/kg at 1 meter. (authors)

  6. HTR's role in process heat applications

    International Nuclear Information System (INIS)

    Kuhr, Reiner

    2008-01-01

    Advanced high-temperature nuclear reactors create a number of new opportunities for nuclear process heat applications. These opportunities are based on the high-temperature heat available, smaller reactor sizes, and enhanced safety features that allow siting close to process plants. Major sources of value include the displacement of premium fuels and the elimination of CO 2 emissions from combustion of conventional fuels and their use to produce hydrogen. High value applications include steam production and cogeneration, steam methane reforming, and water splitting. Market entry by advanced high-temperature reactor technology is challenged by the evolution of nuclear licensing requirements in countries targeted for early applications, by the development of a customer base not familiar with nuclear technology and related issues, by convergence of oil industry and nuclear industry risk management, by development of public and government policy support, by resolution of nuclear waste and proliferation concerns, and by the development of new business entities and business models to support commercialization. New HTR designs may see a larger opportunity in process heat niche applications than in power given competition from larger advanced light water reactors. Technology development is required in many areas to enable these new applications, including the commercialization of new heat exchangers capable of operating at high temperatures and pressures, convective process reactors and suitable catalysts, water splitting system and component designs, and other process-side requirements. Key forces that will shape these markets include future fuel availability and pricing, implementation and monetization of CO 2 emission limits, and the formation of international energy and environmental policy that will support initiatives to provide the nuclear licensing frameworks and risk distribution needed to support private investment. This paper was developed based on a plenary

  7. Introduction of HTR-PM Operation and Fuel Management System

    International Nuclear Information System (INIS)

    Liu Fucheng; Luo Yong; Gao Qiang

    2014-01-01

    There is a big difference between High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) and PWR in operation mode. HTR-PM is a continually refuelled reactor, and the operation and fuel management of it, which affect each other, are inseparable. Therefore, the analysis of HTR-PM fuel management needs to be carried out “in real time”. HTR-PM operation and fuel management system is developed for on-power refuelling mode of HTR-PM. The system, which calculates the core neutron flux and power distribution, taking high-temperature reactor physics analysis software-VSOP as a basic tool, can track and predict the core state online, and it has the ability to restructure core power distribution online, making use of ex-core detectors to correct and check tracking calculation. Based on the ability to track and predict, it can compute the core parameters to provide support for the operation of the reactor. It can also predict the operation parameters of the reactor to provide reference information for the fuel management.The contents of this paper include the development purposes, architecture, the main function modules, running process, and the idea of how to use the system to carry out HTR-PM fuel management. (author)

  8. The future of HTR development and market chances

    International Nuclear Information System (INIS)

    Baust, E.; Weisbrodt, I.

    1989-01-01

    In more than thirty years of development, the pebble bed high-temperature reactor has been brought to the threshold of commercial maturity. On the basis of the experience accumulated with the 15 MW AVR reactor and the THTR-300, unit sizes tailored to demand (HTR-500, modular HTR, GHR-10) will be developed for the electricity and heat markets of the future. The high-temperature reactor is a meaningful supplement to the proven line of light-water reactors and is particularly suitable for being exported to developing countries and industrial threshold countries because of its special technical and inherent safeguards properties. There is broad worldwide interest in the HTR, as is evidenced by several existing agreements on cooperation. It is for this reason that market chances are believed to exist for the HTR after the expected revival of the nuclear power market. ABB and Siemens therefore have decided to develop and market the HTR jointly in the future as a matter of long term strategy by working through a joint subsidiary, HTR-GmbH. (orig.) [de

  9. Why HTR/VHTR? A European point of view

    International Nuclear Information System (INIS)

    Basini, V.; Bogusch, E.; Breuil, E.; Buckthorpe, D.; Chauvet, V.; Ftitterer, M.; Van Heek, A.; Hittner, D.; Von Lensa, W.; Pirson, J.; Verrier, D.

    2008-01-01

    The (European) High Temperature Reactor Technology Network (HTR-TN) was created in 2000 by the main industrial and Research actors of nuclear energy in Europe for elaborating a strategy for developing advanced HTR technology towards industrial application and for taking initiatives for implementing this strategy, most particularly through the Euratom funded R and D programmes. HTR-TN members are convinced that the main market push for industrial deployment of a new generation of HTR will not come from utility needs for electricity generation, but from industrial process heat needs: even if HTR can be considered for satisfying particular niches of the electricity market, there will not be any incentive for utilities already experienced in the exploitation of large LWR to take the risk of a significant technology change, when no evident competitive edge would result from it. On the contrary, HTR is the sole nuclear system that can address heat needs of a large number of industrial processes that require a higher temperature than the temperature provided by all other types of industrial reactors. The possibility for HTR to address the industrial process heat market is a strong asset, as it opens to HTR a large market which is presently looking for solutions to reduce drastically CO 2 emissions, but at the same time it is a huge challenge: industrial exploitation of nuclear energy has been for the time being focused on electricity generation for which user requirements are relatively uniform. The versatility of process heat needs in terms of power, temperature, reliability, etc. will require a much larger flexibility of the nuclear heat source, which is not usual for nuclear industry, looking for competitiveness through standardisation. Therefore HTR-TN considers that the top priority innovation for HTR present development should not be missed: it is to demonstrate at an industrial scale the technical, industrial and economical feasibility of the coupling of a HTR with

  10. Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

    International Nuclear Information System (INIS)

    Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

    1987-01-01

    Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

  11. Hydraulic Model Tests on Modified Wave Dragon

    DEFF Research Database (Denmark)

    Hald, Tue; Lynggaard, Jakob

    A floating model of the Wave Dragon (WD) was built in autumn 1998 by the Danish Maritime Institute in scale 1:50, see Sørensen and Friis-Madsen (1999) for reference. This model was subjected to a series of model tests and subsequent modifications at Aalborg University and in the following...... are found in Hald and Lynggaard (2001). Model tests and reconstruction are carried out during the phase 3 project: ”Wave Dragon. Reconstruction of an existing model in scale 1:50 and sequentiel tests of changes to the model geometry and mass distribution parameters” sponsored by the Danish Energy Agency...

  12. Verification of DRAGON: the NXT tracking module

    International Nuclear Information System (INIS)

    Zkiek, A.; Marleau, G.

    2007-01-01

    The version of DRAGON-IST that has been verified for the calculation of the incremental cross sections associated with CANDU reactivity devices is version 3.04Bb that was released in 2001. Since then, various improvements were implemented in the code including the NXT: module that can track assemblies of clusters in 2-D and 3-D geometries. Here we will discuss the verification plan for the NXT: module of DRAGON, illustrate the verification procedure we selected and present our verification results. (author)

  13. Dragons' Den: promoting healthcare research and innovation.

    Science.gov (United States)

    Mazhindu, Deborah; Gregory, Siobhan

    2015-07-01

    The changing health and social care landscape, and, in particular, the financial challenges affecting the NHS, can present difficulties for staff looking for funding to support innovation and new ways of working. One method of competitive tendering that is becoming more accepted as a way of allocating funds, encouraging staff engagement and developing innovation for research is a format based the BBC television series, Dragons' Den. This article describes how Hounslow and Richmond Community Healthcare NHS Trust, London, has developed a 'Dragons' Den initiative' of annual competitive research funding allocation to ensure that some of the most dynamic practice in the trust is captured.

  14. The second Euratom sponsored 9000C HTR fuel irradiation experiment in the HFR Petten Project E 96.02: Pt.2. Post-irradiation examination

    International Nuclear Information System (INIS)

    Roettger, R.; Bueger, J. de; Schoots, T.

    1977-01-01

    A large variety of HTR fuel specimens, loose coated particles, coupons and compacts provided by Belgonucleaire, the Dragon Project and the KFA Juelich have been irradiated in the HFR at Petten at about 900 0 C up to a maximum fast neutron fluence of about 7x10 21 cm -2 (EDN) as a Euratom sponsored experiment. The maximum burn-ups were between 11 and 18.5% FIMA. The results of the post-irradiation examinations, comprising visual inspection, dimensional measurements, microradiography, metallography, and burn-up determinations are presented in this part 2 of the final report. The examinations have shown that the endurance limit of most of the tested fuel varieties is beyond the reached irradiation values

  15. Criticality calculations of the HTR-10 pebble-bed reactor with SCALE6/CSAS6 and MCNP5

    International Nuclear Information System (INIS)

    Wang, Meng-Jen; Sheu, Rong-Jiun; Peir, Jinn-Jer; Liang, Jenq-Horng

    2014-01-01

    CSAS6 computation time in the high-fidelity HTR-10 model because of the presence of graphite balls and the complex modeling of peripheral reflectors

  16. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  17. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  18. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Mazaira, Leorlen Y.R.; Dominguez, Dany S.; Hernandez, Carlos R.G.

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  19. Oscillator measurements of the reactivity changes resulting from the irradiation of low enrichment particulate fuel in the Dragon reactor

    International Nuclear Information System (INIS)

    Burbidge, B.L.H.; Franklin, B.M.; Small, V.G.

    1983-01-01

    This Report describes a series of experiments carried out as a joint UKAEA/CEA/DRAGON project to determine the reactivity changes of low-enrichment particulate fuel samples following their irradiation in the DRAGON reactor to various levels up to approximately 60,000 MWD/Te. The samples are described, together with the method of measurement of reactivity in the Winfrith reactor HECTOR, which was an extension of the well-known Oscillator Technique to yield simultaneously overall reactivity changes and changes in macroscopic absorption cross-sections. Measurements were carried out at room temperature in two reactor spectra; a thermal spectrum and one typical of an HTR type reactor. The resultant reactivity changes are presented together with the relevant sample burn-ups as determined by #betta#-scanning methods and, in some cases, by rigorous chemical analysis. The results of supporting measurements are also reported, carried out to characterise the neutron spectra in which the oscillator measurements were made and to determine the neutron flux distributions in the HECTOR reactor. (author)

  20. Advances in HTR fuel matrix technology

    International Nuclear Information System (INIS)

    Voice, E.H.; Sturge, D.W.

    1974-02-01

    Progress in the materials and technology of matrix consolidation in recent years is summarised, noting especially the development of an improved resin and the introduction of a new graphite powder. An earlier irradiation programme, the Matrix Test Series, is recalled and the fabrication of the most recent experiment, the directly-cooled homogeneous Met. VI, is described. (author)

  1. CFD investigating the air ingress accident for a HTGR simulation of graphite corrosion oxidation

    International Nuclear Information System (INIS)

    Ferng, Y.M.; Chi, C.W.

    2012-01-01

    Highlights: ► A CFD model is proposed to investigate graphite oxidation corrosion in the HTR-10. ► A postulated air ingress accident is assumed in this paper. ► Air ingress flowrate is the predicted result, instead of the preset one. ► O 2 would react with graphite on pebble surface, causing the graphite corrosion. ► No fuel exposure is predicted to be occurred under the air ingress accident. - Abstract: Through a compressible multi-component CFD model, this paper investigates the characteristics of graphite oxidation corrosion in the HTR-10 core under the postulated accident of gas duct rupture. In this accident, air in the steam generator cavity would enter into the core after pressure equilibrium is achieved between the core and the cavity, which is also called as the air ingress accident. Oxygen in the air would react with graphite on pebble surface, subsequently resulting in oxidation corrosion and challenging fuel integrity. In this paper, characteristics of graphite oxidation corrosion during the air ingress accident can be reasonably captured, including distributions of graphite corrosion amount on the different cross-sections, time histories of local corrosion amount at the monitoring points and overall corrosion amount in the core, respectively. Based on the transient simulation results, the corrosion pattern and its corrosion rate would approach to the steady-state conditions as the accident continuously progresses. The total amount of graphite corrosion during a 3-day accident time is predicted to be about 31 kg with the predicted asymptotic corrosion rate. This predicted value is less than that from the previous work of Gao and Shi.

  2. Weeping dragon, a unique ornamenal citrus

    Science.gov (United States)

    ‘Weeping Dragon’ is a new ornamental citrus cultivar developed by intercrossing of two unusual and unique citrus types, Poncirus trifoliata cultivated variety (cv.) Flying Dragon, and Citrus sinensis cv. ‘Cipo’. This new hybrid cultivar combines strongly contorted and weeping growth traits in a smal...

  3. Safety assessment for Dragon fuel element production

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1963-11-01

    This report shall be the Safety Assessment covering the manufacture of the First Charge of Fuel and Fuel Elements for the Dragon Reactor Experiment. It is issued in two parts, of which Part I is descriptive and Part II gives the Hazards Analysis, the Operating Limitations, the Standing Orders and the Emergency Drill. (author)

  4. Dragon Boat Festival (Dyun Ngh Jit).

    Science.gov (United States)

    Wu, Julia; Quan, Ella Y.

    This bilingual-bicultural reader in Cantonese and English is intended for elementary school children in a bilingual education setting. Pen-and-ink drawings illustrate the story of the traditional dragon boat festival. Each page of the text is written in Chinese characters, Romanized form, and in English. (NCR)

  5. Nuclear astrophysics with DRAGON at ISAC

    International Nuclear Information System (INIS)

    D'Auria, J.M.

    2003-01-01

    A new facility, DRAGON, designed specifically to measure radiative proton and alpha capture reaction rates using short-lived, radioactive beams is almost installed at the new ISAC accelerated radioactive beam facility. A description of the planned experimental program, status of the installation (as of July 2001), results from commissioning studies, and the planned schedule are provided in this report. (orig.)

  6. Ultrasonographic anatomy of bearded dragons (Pogona vitticeps).

    Science.gov (United States)

    Bucy, Daniel S; Guzman, David Sanchez-Migallon; Zwingenberger, Allison L

    2015-04-15

    To determine which organs can be reliably visualized ultrasonographically in bearded dragons (Pogona vitticeps), describe their normal ultrasonographic appearance, and describe an ultrasonographic technique for use with this species. Cross-sectional study. 14 healthy bearded dragons (6 females and 8 males). Bearded dragons were manually restrained in dorsal and sternal recumbency, and coelomic organs were evaluated by use of linear 7- to 15-MHz and microconvex 5- to 8-MHz transducers. Visibility, size, echogenicity, and ultrasound transducer position were assessed for each organ. Coelomic ultrasonography with both microconvex and linear ultrasound transducers allowed for visualization of the heart, pleural surface of the lungs, liver, caudal vena cava, aorta, ventral abdominal vein, gallbladder, fat bodies, gastric fundus, cecum, colon, cloaca, kidneys, and testes or ovaries in all animals. The pylorus was visualized in 12 of 14 animals. The small intestinal loops were visualized in 12 of 14 animals with the linear transducer, but could not be reliably identified with the microconvex transducer. The hemipenes were visualized in 7 of 8 males. The adrenal glands and spleen were not identified in any animal. Anechoic free coelomic fluid was present in 11 of 14 animals. Heart width, heart length, ventricular wall thickness, gastric fundus wall thickness, and height of the caudal poles of the kidneys were positively associated with body weight. Testis width was negatively associated with body weight in males. Results indicated coelomic ultrasonography is a potentially valuable imaging modality for assessment of most organs in bearded dragons and can be performed in unsedated animals.

  7. Experimental Overtopping Investigation for the Wave Dragon

    DEFF Research Database (Denmark)

    Borgarino, Bruno; Kofoed, Jens Peter; Tedd, James

    The present report displays the results from overtopping tests carried on the 1:51.8 Wave Dragon model in September 2007. This tests have been carried on by Bruno Borgarino, James Tedd and Jens Peter Kofoed in the wave tank facilities of Aalborg University. The objective was to provide an updated...

  8. Chinese Dragons in an American Science Unit

    Science.gov (United States)

    Lew, Lee Yuen; McLure, John W.

    2005-01-01

    Can art and science find a happy home in the same unit? We think the answer is yes, if the central problem interests the students and allows them to try out multiple abilities. The sixth-grade unit described in this article, which we called "The Dragon Project," grew mainly from two roots, a study of ancient China and a later probe into…

  9. DRAGON, Reactor Cell Calculation System with Burnup

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the

  10. Aerobic salivary bacteria in wild and captive Komodo dragons.

    Science.gov (United States)

    Montgomery, Joel M; Gillespie, Don; Sastrawan, Putra; Fredeking, Terry M; Stewart, George L

    2002-07-01

    During the months of November 1996, August 1997, and March 1998, saliva and plasma samples were collected for isolation of aerobic bacteria from 26 wild and 13 captive Komodo dragons (Varanus komodoensis). Twenty-eight Gram-negative and 29 Gram-positive species of bacteria were isolated from the saliva of the 39 Komodo dragons. A greater number of wild than captive dragons were positive for both Gram-negative and Gram-positive bacteria. The average number of bacterial species within the saliva of wild dragons was 46% greater than for captive dragons. While Escherichia coli was the most common bacterium isolated from the saliva of wild dragons, this species was not present in captive dragons. The most common bacteria isolated from the saliva of captive dragons were Staphylococcus capitis and Staphylococcus capitis and Staphylococcus caseolyticus, neither of which were found in wild dragons. High mortality was seen among mice injected with saliva from wild dragons and the only bacterium isolated from the blood of dying mice was Pasteurella multocida. A competitive inhibition enzyme-linked immunosorbent assay revealed the presence of anti-Pasteurella antibody in the plasma of Komodo dragons. Four species of bacteria isolated from dragon saliva showed resistance to one or more of 16 antimicrobics tested. The wide variety of bacteria demonstrated in the saliva of the Komodo dragon in this study, at least one species of which was highly lethal in mice and 54 species of which are known pathogens, support the observation that wounds inflicted by this animal are often associated with sepsis and subsequent bacteremia in prey animals.

  11. HTR-Proteus Pebble Bed Experimental Program Cores 5,6,7,&8: Columnar Hexagonal Point-on-Point Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Snoj, Luka [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lengar, Igor [Idaho National Lab. (INL), Idaho Falls, ID (United States); Koberl, Oliver [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  12. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  13. Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

    International Nuclear Information System (INIS)

    Van Heek, A.I.

    1996-06-01

    Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

  14. State of the art in HTR engineering and design

    International Nuclear Information System (INIS)

    Baust, E.

    1984-11-01

    The high-temperature reactor is an universally applicable energy source on the electricity and heat market, providing energy safely, compatible with the environment, and economically. The startup of the THTR-300, which will commence power generation in spring 1985, and the good results of the preparatory tests and studies for the subsequent plant, the HTR-500, created the required preconditions for the placing of an order to commence work to realize the first planning stage of the HTR-500. The order is expected to be placed within short. BBC/HRB has gained a reputation worldwide as the leading manufacturer of HTR plants. BBC/HRB has the know-how to offer HTR plants of various size over the entire capacity range between 100 and 600 MWe, or as twin-type plants up to 1200 MWe, their design being based on the THTR-300 reference plant. The HTR is an uncomplicated reactor system offering many advantages in terms of operation and safety. This reactor type therefore is the system of choice for energy generation for short-range energy supply. The system also is of interest as an export item, and hence is of significance to the economy and to employment policy. (orig.) [de

  15. Multicavity PCPVs for HTR and GCFR systems

    International Nuclear Information System (INIS)

    Eadie, D.Mc.D.

    1979-01-01

    There is little extra to report since the presentation of the paper 180/75 Multicavity PCPVs for HTR and GCFR Systems by P.L.T. Morgan and J.N. Bradbury at the International Conference on Experience in the Design, Construction and Operation of Prestressed Concrete Pressure Vessels and Containments for Nuclear Reactors at York, England, in September 1975. The paper presented at the York Conference demonstrated how a particular mode of behaviour could develop in a very local region between the pods and the external wall of a multicavity pressure vessel. Two main points emerge from the paper presented at York - 1. Local analysis for equilibrium of parts of the structure are as important as analysis of the general structural behaviour. With modern computer techniques, in which crack propagation and plasticity may be included, the development of local critical areas can be observed, but the idealisation of the structure has to be sufficiently refined and the cost will be high; 2. Criteria for acceptance of a design must be realistic and must be continually reviewed in the light of the trends of design philosophy. In conclusion, some pictures of model tests demonstrate the physical reality of the mode of failure described in the paper

  16. Microscopic thermal characterization of HTR particle layers

    International Nuclear Information System (INIS)

    Rochais, D.; Le Meur, G.; Basini, V.; Domingues, G.

    2008-01-01

    This paper presents thermal diffusivity measurements of HTR fuel particle pyrolytic carbon layers at room temperature. The photoreflectance microscopy (PM) technique is used to characterize particle layers at a microscopic scale. Nevertheless, buffer layer needs a particular analysis due to its porous structure. Indeed, measurements by PM on this material only permit to obtain the thermal diffusivity of the solid skeleton, whose homogeneous zones surface does not exceed 100 μm 2 . These characteristics make, on the one hand, delicate the use of PM, and on the other hand, require the use of a numerical homogenization technique. This model takes into account the properties of gas confined in the pores, to simulate the conduction heat flux traveling through the layer in relation with its microstructure and to estimate an effective thermal conductivity of the entire layer. This approach is validated by infrared microscopy measurement of the effective thermal diffusivity of the especially elaborated thicker buffer layer. Last, the first tests to characterize the silicon carbide layer are presented

  17. Results and future programme of HTR's study

    International Nuclear Information System (INIS)

    Mursid Djokolelono; Soedyartomo Soentono

    1990-01-01

    Study on the application of HTRs for the enhanced oil recovery in the Duri oil field (Sumatra, Indonesia) was performed in 1986/1987. The economic and technological advantages over crude burning option were identified. Crude oil prices, HTR capital costs, discount rates and company's income structure represented dominant parameters. Further sensitivity calculations on important economic parameters were obtained to reflect the condition of 1988. This nuclear option was also incorporated in the energy planning study for the whole of Indonesia using the MARKAL model, and resulted in the conditions of its applicability. The scenarios chosen in this MARKAL study were high and low GDP growth rate, whereas the criteria chosen were the minimum cost with and without a predetermined policy of reduced domestic use of oil. In the high scenario the HTRs as well as the natural gas options could not compete against the low cost boilers with crude-oil fuel. But in the case of reduced domestic oil use the HTRs came out to supplement the crudeburning boilers starting in the sixth five year plan (1994-999), even earlier than the natural gas option. The authors further discuss the industrial environment, in relation to the regional development, the possible local participation, as well as the plan to materialize the merits of this novel application. (author)

  18. Results and future programme of HTR's study

    Energy Technology Data Exchange (ETDEWEB)

    Djokolelono, Mursid; Soentono, Soedyartomo [National Atomic Energy Agency (Indonesia)

    1990-07-01

    Study on the application of HTRs for the enhanced oil recovery in the Duri oil field (Sumatra, Indonesia) was performed in 1986/1987. The economic and technological advantages over crude burning option were identified. Crude oil prices, HTR capital costs, discount rates and company's income structure represented dominant parameters. Further sensitivity calculations on important economic parameters were obtained to reflect the condition of 1988. This nuclear option was also incorporated in the energy planning study for the whole of Indonesia using the MARKAL model, and resulted in the conditions of its applicability. The scenarios chosen in this MARKAL study were high and low GDP growth rate, whereas the criteria chosen were the minimum cost with and without a predetermined policy of reduced domestic use of oil. In the high scenario the HTRs as well as the natural gas options could not compete against the low cost boilers with crude-oil fuel. But in the case of reduced domestic oil use the HTRs came out to supplement the crudeburning boilers starting in the sixth five year plan (1994-999), even earlier than the natural gas option. The authors further discuss the industrial environment, in relation to the regional development, the possible local participation, as well as the plan to materialize the merits of this novel application. (author)

  19. Status of the HTR programme in France

    International Nuclear Information System (INIS)

    Ballot, B.; Gauthier, J.C.; Hittner, D.; Lebrun, J.Ph.; Lecomte, M.; Carre, F.; Delbecq, J.M.

    2007-01-01

    AREVA is convinced that HTR (High Temperature Reactor) is not in competition with large LWRs for electricity generation, and that its actual added value is its potential for addressing cogeneration and industrial process heat production. Therefore AREVA launched in 2004 the ANTARES programme for a pre-conceptual design study, with the support of EDF and together with a large research and development programme needed for the design in close collaboration with Cea. The pre-conceptual phase was finalized end of 2006. The specific feature of AREVA's concept, which distinguishes it from other ones, is its indirect cycle design powering a combined cycle power plant. Several reasons support this design choice, one of the most important being the design flexibility to adapt readily to combined heat and power applications, with a standardised nuclear heat source as independent as possible of the versatile process heat applications with which it is coupled. Standardisation should expedite licensing. In view of the volatility of the costs of fossil fuels, AREVA's choice brings to the large industrial heat applications the fuel cost predictability of nuclear fuel with the efficiency of a high temperature heat source free of greenhouse gases emissions. The reactor module produces 600 MWth which can be split into the required process heat, the remaining power driving an adapted prorated electric plant. Depending on the process heat temperature and power needs, up to 80 % of the nuclear heat is converted into useful energy

  20. Dragon (RGMb) induces oxaliplatin resistance in colon cancer cells.

    Science.gov (United States)

    Shi, Ying; Huang, Xiao-Xiao; Chen, Guo-Bin; Wang, Ying; Zhi, Qiang; Liu, Yuan-Sheng; Wu, Xiao-Ling; Wang, Li-Fen; Yang, Bing; Xiao, Chuan-Xing; Xing, Hui-Qin; Ren, Jian-Lin; Xia, Yin; Guleng, Bayasi

    2016-07-26

    Colorectal cancer (CRC) is one of the most commonly diagnosed cancers and a major cause of cancer mortality. Chemotherapy resistance remains a major challenge for treating advanced CRC. Therefore, the identification of targets that induce drug resistance is a priority for the development of novel agents to overcome resistance. Dragon (also known as RGMb) is a member of the repulsive guidance molecule (RGM) family. We previously showed that Dragon expression increases with CRC progression in human patients. In the present study, we found that Dragon inhibited apoptosis and increased viability of CMT93 and HCT116 cells in the presence of oxaliplatin. Dragon induced resistance of xenograft tumor to oxaliplatinin treatment in mice. Mechanistically, Dragon inhibited oxaliplatin-induced JNK and p38 MAPK activation, and caspase-3 and PARP cleavages. Our results indicate that Dragon may be a novel target that induces drug resistance in CRC.

  1. Phonon scattering in graphite

    International Nuclear Information System (INIS)

    Wagner, P.

    1976-04-01

    Effects on graphite thermal conductivities due to controlled alterations of the graphite structure by impurity addition, porosity, and neutron irradiation are shown to be consistent with the phonon-scattering formulation 1/l = Σ/sub i equals 1/sup/n/ 1/l/sub i/. Observed temperature effects on these doped and irradiated graphites are also explained by this mechanism

  2. Periodic safety review of the HTR-10 safety analysis

    International Nuclear Information System (INIS)

    Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

    2015-01-01

    Designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first modular High Temperature Gas-cooled Reactor (HTGR) in China. According to the nuclear safety regulations of China, the periodic safety review (PSR) of the HTR-10 was initiated by INET after approved by the National Nuclear Safety Administration (NNSA) of China. Safety analysis of the HTR-10 is one of the key safety factors of the PSR. In this paper, the main contents in the review of safety analysis are summarized; meanwhile, the internal evaluation on the review results is presented by INET. (authors)

  3. Overview of Japanese seismic research program for HTR

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1978-07-01

    In order to obtain the license for construction and operation of HTR developed and introduced into Japan, it is necessary to assure integrity of reactor structures and the capability of reactor shutdown and maintain safety shutdown for the seismic design condition. Because Japanese land is located in relatively high seismacity zone, when an excessive earthquake would occur, the public and plant personnel should be protected from radiation hazard. For the above reason, many efforts of seismic research and development for HTR have been made at institutes and companies in Japan. In the paper, descriptions are: (1) Present status of development and construction plans of HTR, (2) guideline of aseismic design, (3) need of aseismic research, (4) present status of research and development, (5) future plan. (auth.)

  4. HTR-E project. High-temperature components and systems

    International Nuclear Information System (INIS)

    Breuil, E.; Exner, R.

    2002-01-01

    The HTR-E European project (four years project) is proposed for the 5th Framework Programme and concerns the technical developments needed for the innovative components of a modern HTR with a direct cycle. These components have been selected with reference to the present projects (GT-MHR, PBMR): (1) the helium turbine, the recuperator heat exchanger, the electro-magnetic bearings and the helium rotating seal; (2) the tribology. Sliding innovative components in helium environment are particularly concerned. (3) the helium purification system. Recommendations on impurities contents have to be provided in accordance with the materials proposed for the innovative components. The main outcomes expected from the HTR-E project are the design recommendations and identification of further R and D needs for these components. This will be based: (1) on experience feedback from European past helium test loops and reactors; (2) on design studies, thermal-hydraulic and structural analyses; (3) and on experimental tests

  5. Gas reactor international cooperative program. HTR-synfuel application assessment

    International Nuclear Information System (INIS)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H 2 +CO 2 ) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000 0 F steam is generated at the industrial user sites. The products of methanation (CH 4 + H 2 O) are piped back to the reformer at the central station HTR

  6. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  7. Risk assessment of small-sized HTR with pebble-bed core

    International Nuclear Information System (INIS)

    Kroeger, W.; Mertens, J.; Wolters, J.

    1987-01-01

    Two recent concepts of small-sized HTR's (HTR-Modul and HTR-100) were analysed regarding their safety concepts and risk protection. In neither case do core cooling accidents contribute to the risk because of the low induced core temperatures. Water ingress accidents dominate the risk in both cases by detaching deposited fission products which can be released into the environment. For these accident sequences no early fatalities and practically no lethal case of cancer were computed. Both HTR concepts include adequate precautionary measures and an infinitely small risk according to the usual standards. The safety concepts make express use of the specific inherent safety features of pebble-bed HTR's. (orig.)

  8. Burnable poison management in a HTR

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, J

    1971-09-21

    It is the purpose with this paper to describe the state-of-the-art of burnable poison investigations made within the Dragon Project and to give the results of a number of calculations, which show that it is possible to control the large initial surplus reactivity of the first core and the radial power distribution with two types of burnable poison sticks with Gadolinium (one type of stick to be used in the inner core region, the other in the outer core region), where the poison will burn away so that keff always stays around the desired value 1.03, and with the radial form-factor not exceeding 1.20. The calculations made for this paper are not too accurate, especially the chosen timestep for calculating the burn-up of the burnable poison stick proved to be too large. Nevertheless, the calculations are good enough to draw the above mentioned conclusions, although they have not given the concentration of Gadolinium to be used in the burnable poison sticks very accurately.

  9. HTR process heat applications, status of technology and economical potential

    International Nuclear Information System (INIS)

    Barnet, H.

    1997-01-01

    The technical and industrial feasibility of the production of high temperature heat from nuclear fuel is presented. The technical feasibility of high temperature heat consuming processes is reviewed and assessed. The conclusion is drawn that the next technological step for pilot plant scale demonstration is the nuclear heated steam reforming process. The economical potential of HTR process heat applications is reviewed: It is directly coupled to the economical competitiveness of HTR electricity production. Recently made statements and pre-conditions on the economic competitiveness in comparison to world market coal are reported. (author). 8 figs

  10. Relevant safety issues in designing the HTR-10 reactor

    International Nuclear Information System (INIS)

    Sun Yuliang; Xu Yuanghui

    2001-01-01

    The HTR-10 is a 10 MWth pebble bed high temperature gas cooled reactor being constructed as a research facility at the Institute of Nuclear Energy Technology. This paper discusses design issues of the HTR-10 which are related to safety. It addresses the safety criteria used in the development and assessment of the design, the safety important systems, and the safety classification of components. It also summarises the results of safety analysis, including the approach used for the radioactive source term, as well as the approach to containment design. (author)

  11. Energy analysis of control rod drive mechanism in HTR-10

    International Nuclear Information System (INIS)

    Bo Hanliang; Wu Yuanqiang

    2000-01-01

    This paper presents a theoretical model for the control rod drive mechanism for the 10 MW High Temperature Gas Cooled Reactor (HTR-10) and analyzes accidents which may occur in the drive mechanism, for example, chain break, coupling damage and other damage scenarios. The results show that the matching problem between buffer capability and coupling strength is the main reason for coupling damage; increased temperatures would reduce eddy damping and cause a mismatch between buffer capability and coupling strength; and the displacement of the buffer spring will affect the coupling force. The results provide a theoretical basis for the design of the control rod drive mechanism for HTR-10

  12. Engineering and licensing progress of the HTR-Module

    Energy Technology Data Exchange (ETDEWEB)

    Weisbrodt, I A

    1988-07-01

    This report deals not only with the latest status of Siemens/Interatom's HTR-Module but also reflects the latest engineering and licensing progress of the HTR-Module against the background of the specified design requirements and of the discussions on passively safe reactors. Therefore, I intend to report also about two examples of the accident analysis - one design basis accident, i.e. the leak-before-break of the reactor pressure vessel and one beyond design accident, i. e. massive water ingress.

  13. Engineering and licensing progress of the HTR-Module

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1988-01-01

    This report deals not only with the latest status of Siemens/Interatom's HTR-Module but also reflects the latest engineering and licensing progress of the HTR-Module against the background of the specified design requirements and of the discussions on passively safe reactors. Therefore, I intend to report also about two examples of the accident analysis - one design basis accident, i.e. the leak-before-break of the reactor pressure vessel and one beyond design accident, i. e. massive water ingress

  14. Overview of Japanese seismic research program for HTR

    International Nuclear Information System (INIS)

    Ikushima, T.

    1978-01-01

    In order to obtain the license for construction and operation of HTR developed in and/or introduced into Japan, it is necessary to insure the integrity of reactor structures and the capability of reactor shutdown and the maintenance of safety shutdown for the seismic design condition. Because Japan is located in relatively high seismicity zone, even when an excessive earthquake would occur, the public and plant personnel should be protected from radiation hazard. The report describes the following: (1) present status of development and construction plan of HTR, (2) guideline of aseismic design, (3) need of aseismic research, (4) present status of research and development, and (5) future plans

  15. Symbiosis of near breeder HTR's with hybrid fusion reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1978-07-01

    In this contribution to INFCE a symbiotic fusion/fission reactor system, consisting of a hybrid beam-driven micro-explosion fusion reactor (HMER) and associated high-temperature gas-cooled reactors (HTR) with a coupled fuel cycle, is proposed. This system is similar to the well known Fast Breeder/Near Breeder HTR symbiosis except that the fast fission breeder - running on the U/Pu-cycle in the core and the axial blankets and breeding the surplus fissile material as U-233 in its radial thorium metal or thorium oxide blankets - is replaced by a hybrid micro-explosion DT fusion reactor

  16. IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Burchell, T.D.; Clark, R.E.H.; Stephens, J.A.; Eto, M.; Haag, G.; Hacker, P.; Neighbour, G.B.; Janev, R.K.; Wickham, A.J.

    2000-02-01

    This report describes an IAEA database containing data on the properties of irradiated nuclear graphites. Development and implementation of the graphite database followed initial discussions at an IAEA Specialists' Meeting held in September 1995. The design of the database is based upon developments at the University of Bath (United Kingdom), work which the UK Health and Safety Executive initially supported. The database content and data management policies were determined during two IAEA Consultants' Meetings of nuclear reactor graphite specialists held in 1998 and 1999. The graphite data are relevant to the construction and safety case developments required for new and existing HTR nuclear power plants, and to the development of safety cases for continued operation of existing plants. The database design provides a flexible structure for data archiving and retrieval and employs Microsoft Access 97. An instruction manual is provided within this document for new users, including installation instructions for the database on personal computers running Windows 95/NT 4.0 or higher versions. The data management policies and associated responsibilities are contained in the database Working Arrangement which is included as an Appendix to this report. (author)

  17. Transient behaviour of small HTR for cogeneration

    International Nuclear Information System (INIS)

    Verkerk, E.C.; Van Heek, A.I.

    2000-01-01

    The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of a nuclear cogeneration plant. The ACACIA system provides 14 MWe electricity combined with 17 t/h of high temperature steam (220 deg C, 10 bar) with a pebble-bed high temperature reactor directly coupled with a helium compressor and a helium turbine. The design of this small CHP unit that is used for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. Thermal hydraulic and reactor physics analyses show favourable control characteristics during normal operation and a benign response to loss of helium coolant and loss of flow conditions. Throughout the response on these highly infrequent conditions, ample margin exists between the highest fuel temperatures and the temperature above which fuel degradation will occur. To come to quantitative statements about the ACACIA transient behaviour, a calculational coupling between the high temperature reactor core analysis code package PANTHER/DIREKT and the thermal hydraulic code RELAP5 for the energy conversion system has been made. This coupling offers a more realistic simulation of the entire system, since it removes the necessity of forcing boundary conditions on the simulation models at the data transfer points. In this paper, the models used for the dynamic components of the energy conversion system are described, and the results of the calculation for two operational transients in order to demonstrate the effects of the interaction between reactor core and its energy conversion system are shown. Several transient cases that are representative as operational transients for an HTR will be discussed, including one representing a load rejection case that shows the functioning of the control system, in particular the bypass valve. Another transient is a load following

  18. DRAGON - 8U Nanosatellite Orbital Deployer

    Science.gov (United States)

    Dobrowolski, Marcin; Grygorczuk, Jerzy; Kedziora, Bartosz; Tokarz, Marta; Borys, Maciej

    2014-01-01

    The Space Research Centre of the Polish Academy of Sciences (SRC PAS) together with Astronika company have developed an Orbital Deployer called DRAGON for ejection of the Polish scientific nanosatellite BRITE-PL Heweliusz (Fig. 1). The device has three unique mechanisms including an adopted and scaled lock and release mechanism from the ESA Rosetta mission MUPUS instrument. This paper discusses major design restrictions of the deployer, unique design features, and lessons learned from development through testing.

  19. A new impetus for developing industrial process heat applications of HTR in europe - HTR2008-58259

    International Nuclear Information System (INIS)

    Hittner, D.; De Groot, S.; Griffay, G.; Yvon, P.; Pienkowski, L.; Ruer, J.; Angulo, C.; Laquaniello, G.

    2008-01-01

    Due to its high operating temperature (up to 850 deg. C with present technologies, possibly higher in the longer term), and its power range (a few hundred MW), the modular HTR could address a larger scope of industrial process heat needs than other present nuclear systems. Even if HTR can contribute to competitive electricity generation, this potential for industrial heat applications is the main incentive for developing this type of reactor, as it could open to nuclear energy a large non-electricity market. However several issues must be addressed and solved successfully for HTR to actually enter the market of industrial process heat: 1) as an absolute prerequisite, to develop a strategic alliance of nuclear industry and R and D with process heat user industries. 2) to solve some key technical issues, as for instance the design of a reactor and of a coupling system flexible enough to reconcile a single reactor design with multiple applications and versatile requirements for the heat source, and the development of special adaptations of the application processes or even of new processes to fit with the assets and constraints of HTR heat supply, 3) to solve critical industrial issues such as economic competitiveness, availability and 4) to address the licensing issues raised by the conjunction of nuclear and industrial risks. In line with IAEA initiatives for supporting non-electric applications of nuclear energy and with the orientations of the SET-Plan of the European Commission, the (European) HTR Technology Network (HTR-TN) proposes a new project, together with industrial process heat user partners, to provide a first impetus to the strategic alliance between nuclear and non-nuclear industries. End user requirements will be expressed systematically on the basis of inputs from industrial partners on various types of process heat applications. These requirements will be confronted with the capabilities of the HTR heat source, in order to point out possible

  20. Stick balancing, falls and Dragon-Kings

    Science.gov (United States)

    Cabrera, J. L.; Milton, J. G.

    2012-05-01

    The extent to which the occurrence of falls, the dominant feature of human attempts to balance a stick at their fingertip, can be predicted is examined in the context of the "Dragon-King" hypothesis. For skilled stick balancers, fluctuations in the controlled variable, namely the vertical displacement angle θ, exhibit power law behaviors. When stick balancing is made less stable by either decreasing the length of the stick or by requiring the subject to balance the stick on the surface of a table tennis racket, systematic departures from the power law behaviors are observed in the range of large θ. This observation raises the possibility that the presence of departures from the power law in the large length scale region, possibly Dragon-Kings, may identify situations in which the occurrence of a fall is more imminent. However, whether or not Dragon-Kings are observed, there is a Weibull-type survival function for stick falling. The possibility that increased risk of falling can, at least to some extent, be predicted from fluctuations in the controlled variable before the event occurs has important implications for the development of preventative strategies for the management of phenomena ranging from earthquakes to epileptic seizures to falls in the elderly.

  1. Design of the Wave Dragon Mooring System

    DEFF Research Database (Denmark)

    Parmeggiani, Stefano

    with experimental data, derived from tank tests of the 2nd generation scaled model of the device. In phase 2 further tank testing has been carried out on a novel 3rd generation scaled model to assess the design loads in the mooring system and the extreme response of the device in surge, heave and pitch to extreme...... storm conditions typical of the DanWEC location. The most desirable mooring configuration has also been better defined in terms of horizontal compliance. In phase 3 results from phase 1 shall be used to setup a numerical model for time-domain analysis of the composite system Wave Dragon + moorings......This report is part of the project “Wave Dragon 1.5 MW North Sea Demonstrator”, funded by the Danish Energy Agency under the EUDP program (J.no. 64010-0405). In phase 1 of the project the hydrodynamic characterization of Wave Dragon was carried out through numerical analysis with a model calibrated...

  2. Investigation on wear behavior of graphite baII under different pneumatic conveying environments

    International Nuclear Information System (INIS)

    Chen Zhipeng; Zheng Yanhua; Shi Lei; Yu Suyuan

    2014-01-01

    An experimental platform was built in the Institute of Nuclear and New Energy Technology (INET) to investigate the wear behavior of the graphite ball under the operational condition of the high temperature gas-cooled reactor (HTGR) fuel handling system. In this experimental platform, a series of experiments were carried out under different pneumatic conveying environments with the graphite balls, which were made of the material same as the fuel element matrix graphite (A3) of the 10 MW high temperature gas cooled reactor (HTR-10). The effect of the pneumatic conveying condition on the wear rate of graphite ball has been investigated, and the results include: (1) There is an obvious linear relationship between the wear rate and the feeding velocity of graphite ball elevated in the stainless steel elevating tube, and the wear rate will increase with the increase of the feeding velocity. (2) The wear rate of graphite ball under helium environment is significantly greater than that under air and nitrogen environments, which is caused by the different effects of various gas environments on mechanical properties of graphite. (author)

  3. Analysis of diffusion process and influence factors in the air ingress accident of the HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Yanhua, Zheng, E-mail: zhengyh@mail.tsinghua.edu.cn; Fubing, Chen; Lei, Shi

    2014-05-01

    Air ingress, one of the beyond design basis accidents for high temperature gas-cooled reactors, receives high attention during the design of the 250 MW pebble-bed modular high temperature gas-cooled reactor (HTR-PM), because it may result in severe consequence including the corrosion of the fuel element and graphite reflector. The diffusion process and the set-up time of the stable natural convection after the double-ended guillotine break of the hot-gas duct are studied in the paper. On the basis of the preliminary design of the HTR-PM and its DLOCA analysis results, the diffusion process, as well as the influence of the core temperature distribution and the length of the hot-gas duct, is studied with the DIFFLOW code, which adopts a one-dimension variable cross-section diffusion model with fixed wall temperature. To preliminarily estimate the influence of chemical reaction between oxygen and graphite, which will change the gas component of the mixture, the diffusion processes between the He/N{sub 2}, He/O{sub 2}, He/CO and He/CO{sub 2} are calculated, respectively. Furthermore, the code has been improved and the varying wall temperature can be simulated. The more accurate analysis is carried out with the changing temperature distribution from the DLOCA calculation. The analysis shows that there is enough time to adopt appropriate mitigation measures to stop the air ingress and the severe consequence of fuel element damage and large release of fission product can be avoided.

  4. Would HTR be suitable for application in the Netherlands?

    International Nuclear Information System (INIS)

    Heek, A.I. van.

    1994-08-01

    The modular HTR may be a reactor type, which would have sufficient societal support to be constructed in the Netherlands. The economic approach would be fundamentally different from that applied in present nuclear technology. In a national research program this is being investigated. (orig.)

  5. On Power Refueling Management of HTR-PM

    International Nuclear Information System (INIS)

    Sun Furui; Luo Yong; Gao Qiang

    2014-01-01

    The refueling management is an important work of nuclear power plant , directly affecting its safety and economy. At present, the ordinary commercial pressurized water reactor (PWR) nuclear power plant has developed more mature in the refueling management, and formed a set of relatively complete system and methods.The High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) has significant differences with the ordinary PWR nuclear power plant in the fuel morphology and the refueling mode. It adopts the spherical fuel element and the on-power refueling. Therefore, the HTR-PM refueling management has its own unique characteristics, but currently there is no mature experience to use for reference across the world. This paper gives a brief introduction to the HTR-PM on power refueling management, including the refueling management system construction, the refueling strategy, the fuel element internal transportation,charging and discharging, etc. It aims at finding the befitting HTR-PM refueling management methods in view of its own unique characteristics in order to ensure the orderly development of the refueling management and the refueling safety. (author)

  6. Profiles of facilities used for HTR research and testing

    International Nuclear Information System (INIS)

    1980-05-01

    This report contains a current description of facilities supporting HTR research and development submitted by countries participating in the IWGFR. It has the purpose of providing an overview of the facilities available for use and of the types of experiments that can be conducted therein

  7. Waste heat of HTR power stations for district heating

    International Nuclear Information System (INIS)

    Bonnenberg, H.; Schlenker, H.V.

    1975-01-01

    The market situation, the applied techniques, and the transport, for district heating in combination with HTR plants are considered. Analysis of the heat market indicates a high demand for heat at temperatures between 100 and 150 0 C in household and industry. This market for district heating can be supplied by heat generated in HTR plants using two methods: (1) the combined heat and power generation in steam cycle plants by extracting steam from the turbine, and (2) the use of waste heat of a closed gas turbine cycle. The heat generation costs of (2) are negligible. The cost for transportation of heat over the average distance between existing plant sites and consumer regions (25 km) are between 10 and 20% of the total heat price, considering the high heat output of nuclear power stations. Comparing the price of heat gained by use of waste heat in HTR plants with that of conventional methods, considerable advantages are indicated for the combined heat and power generation in HTR plants. (author)

  8. Solution of multiple circuits of steam cycle HTR system

    International Nuclear Information System (INIS)

    Li, Fu; Wang, Dengying; Hao, Chen; Zheng, Yanhua

    2014-01-01

    In order to analyze the dynamic operation performance and safety characteristics of the steam cycle high temperature gas cooled reactor (HTR) systems, it is necessary to find the solution of the whole HTR systems with all coupled circuits, including the primary circuit, the secondary circuit, and the residual heat removal system (RHRS). Considering that those circuits have their own individual fluidity and characteristics, some existing code packages for independent circuits themselves have been developed, for example THEMRIX and TINTE code for the primary circuit of the pebble bed reactor, BLAST for once through steam generator. To solve the coupled steam cycle HTR systems, a feasible way is to develop coupling method to integrate these independent code packages. This paper presents several coupling methods, e.g. the equivalent component method between the primary circuit and steam generator which reflect the close coupling relationship, the overlapping domain decomposition method between the primary circuit and the passive RHRS which reflects the loose coupling relationship. Through this way, the whole steam cycle HTR system with multiple circuits can be easily and efficiently solved by integration of several existing code packages. Based on this methodology, a code package TINTE–BLAST–RHRS was developed. Using this code package, some operation performance of HTR–PM was analyzed, such as the start-up process of the plant, and the depressurized loss of forced cooling accident when different number of residual heat removal trains is operated

  9. HTR-2002: Proceedings of the conference on high temperature reactors

    International Nuclear Information System (INIS)

    2002-01-01

    High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing

  10. Experiments in MARIUS on HTR tubular fuel with loose particles

    Energy Technology Data Exchange (ETDEWEB)

    Bosser, R; Langlet, G

    1972-06-15

    The work described on HTR tubular fuel with loose particles is the first part of a program in three points. The cell is the same in the three experiments, only particles in the fuel container are changed. The aim of the experiment is to achieve the buckling in a critical facility. A description of the techniques of measurements, calculations, and results are presented.

  11. Licensing experience of the HTR-10 test reactor

    International Nuclear Information System (INIS)

    Sun, Y.; Xu, Y.

    1996-01-01

    A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

  12. Overtopping Measurements on the Wave Dragon Nissum Bredning Prototype

    DEFF Research Database (Denmark)

    Frigaard, Peter; Kofoed, Jens Peter; Rasmussen, Michael R.

    2004-01-01

    The paper describes the methods used to estimate (calculated from some indirect measurements) the overtopping of the wave energy converter Wave Dragon placed in a real sea environment. The wave energy converter in quistion is the 237-tonne heavy Wave Dragon Nissum Bredning Prototype. Comparisons...

  13. Pigment identification and antioxidant properties of red dragon fruit ...

    African Journals Online (AJOL)

    In the antioxidant properties determination, there were 86.10 mg of total polyphenolic compound in 0.50 g of dried dragon fruit extract using the total polyphenol assay which expresses gallic acid as equivalent. The reducing power assay further confirmed the antioxidant activity present in dragon fruit where the reducing ...

  14. DRAGON analysis of MOX fueled VVER cell benchmarks

    International Nuclear Information System (INIS)

    Marleau, G.; Foissac, F.

    2002-01-01

    The computational unit-cell benchmarks problems for LEU and MOX fueled VVER-1000 ('water-water energetic reactor') have been analyzed using the code DRAGON with ENDF/B-V and ENDF/B-VI based WIMS-AECL cross section libraries. The results obtained were compared with those generated using the SAS2H module of the SCALE-4.3 computational code system and with the code HELIOS. Good agreements between DRAGON and HELIOS were obtained when the ENDF/B-VI based library was considered while the ENDF/B-V DRAGON results were generally closer to those obtained using SAS2H. This study was useful for the verification of the DRAGON code and confirms that HELIOS and DRAGON have a similar behavior when compatible cross sections library are used. (author)

  15. Rudi Stamm'ler contributions and Dragon - 041

    International Nuclear Information System (INIS)

    Roy, R.; Marleau, G.; Hebert, A.

    2010-01-01

    The lattice code DRAGON has been in constant development over the last 25 years. During this period, the DRAGON development team has often been directly influenced by the excellent work of Rudi Stamm'ler. First, his book on reactor physics has inspired a large number of programming and calculation techniques that were implemented in DRAGON. Then, the work of Rudi and his collaborators on the lattice code HELIOS, has also prompted a friendly competition that lead us to continuously improve our code in such a way that it could match the performance achieved by HELIOS. This paper provides a description of some characteristics or technologies implemented in DRAGON that were influenced by the work of Rudi Stamm'ler. It also describes a Candu simulation exercise where the capabilities of the HELIOS and DRAGON codes were combined. (authors)

  16. The R&D of HTGR high temperature helium sampling loop: From HTR-10 to HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Chao, E-mail: fangchao@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); Bao, Xuyin; Yang, Chen; Yang, Yanran; Cao, Jianzhu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China)

    2016-09-15

    A High Temperature Helium Sampling Loop (HTHSL) for studying the transportation (deposition) behavior and total amount of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) and High Temperature Reactor-Pebble bed Modules (HTR-PM) are researched and designed, respectively. Through the optimal design and simulation based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly so that it could be used to study fission products in the primary circuit of HTR-10. On the other hand, an improved DSD is also designed for HTR-PM based on corresponding simulations, which could be used to sample the important nuclei in the high temperature helium from SG. These schemes offer two different methods to obtain the original source term in the high temperature helium, which will provide deeper understanding for the analysis of source terms of HTGR.

  17. A graphite nanoeraser

    DEFF Research Database (Denmark)

    Liu, Ze; Bøggild, Peter; Yang, Jia-rui

    2011-01-01

    We present here a method for cleaning intermediate-size (up to 50 nm) contamination from highly oriented pyrolytic graphite and graphene. Electron-beam-induced deposition of carbonaceous material on graphene and graphite surfaces inside a scanning electron microscope, which is difficult to remove...... by conventional techniques, can be removed by direct mechanical wiping using a graphite nanoeraser, thus drastically reducing the amount of contamination. We discuss potential applications of this cleaning procedure....

  18. Maternal HtrA3 optimizes placental development to influence offspring birth weight and subsequent white fat gain in adulthood.

    Science.gov (United States)

    Li, Ying; Salamonsen, Lois A; Hyett, Jonathan; Costa, Fabricio da Silva; Nie, Guiying

    2017-07-04

    High temperature requirement factor A3 (HtrA3), a member of the HtrA protease family, is highly expressed in the developing placenta, including the maternal decidual cells in both mice and humans. In this study we deleted the HtrA3 gene in the mouse and crossed females carrying zero, one, or two HtrA3-expressing alleles with HtrA3 +/- males to investigate the role of maternal vs fetal HtrA3 in placentation. Although HtrA3 -/- mice were phenotypically normal and fertile, HtrA3 deletion in the mother resulted in intra-uterine growth restriction (IUGR). Disorganization of labyrinthine fetal capillaries was the major placental defect when HtrA3 was absent. The IUGR caused by maternal HtrA3 deletion, albeit being mild, significantly altered offspring growth trajectory long after birth. By 8 months of age, mice born to HtrA3-deficient mothers, independent of their own genotype, were significantly heavier and contained a larger mass of white fat. We further demonstrated that in women serum levels of HtrA3 during early pregnancy were significantly lower in IUGR pregnancies, establishing an association between lower HtrA3 levels and placental insufficiency in the human. This study thus revealed the importance of maternal HtrA3 in optimizing placental development and its long-term impact on the offspring well beyond in utero growth.

  19. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

  20. State of the Art of helium heat exchanger development for future HTR-projects - HTR2008-58146

    International Nuclear Information System (INIS)

    Esch, M.; Juergens, B.; Hurtado, A.; Knoche, D.; Tietsch, W.

    2008-01-01

    In Germany two HTR nuclear power plants had been built and operated, the AVR-15 and the THTR-300. Also various projects for different purposes in a large power range had been developed, The AVR-15, an experimental reactor with a power output of 15 MWel was operated for more than 20 years with excellent results. The THTR-300 was designed as a prototype demonstration plant with 300 MWel and should be the technological basis for the entire future reactor line. The THTR-300 was prematurely shut down and decommissioned because of political reasons. But because of the accompanying comprehensive R and D program and the operation time of about 5 years, the technology was proved and essential operational results were gained. The AVR steam generator was installed above the reactor core. The six THTR heat exchangers were arranged circularly around the reactor core, Both heat exchanger systems have been operated successfully and furthermore acted as a residual heat removal system. The technology knowledge and experience gained on these existing HTR plants is still available at Westinghouse Electric Germany GmbH since Westinghouse is one of the legal successors of the former German HTR companies. As a follow-up project of THTR, the HTR-500 was developed and designed up to the manufacturing stage. For this plant additionally to the 8 steam generators, two residual heat removal heat exchangers were foreseen. These were to be installed in a ring around the reactor core. All these HTRs were designed for the generation of electricity using a steam cycle. Extensive research work has also been done for advanced applications of HTR technology e.g. using a direct cycle within the HHT project or generating process heat within the framework of the PNP project, Because of the critical attitude of the German government to the nuclear power in the past 20 years in Germany there was only a very limited interest in the further development of the HTR technology. As a consequence of the German

  1. Financing models for HTR plants: Co-financing, counter trade, joint ventures

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    Structure and volume of investment cost for HTR nuclear power plants are different in comparison to other types of nuclear power plants. Even if the share of local participation is in comparable order of magnitude to other nuclear power plants, the required technical infrastructure for HTR plants is more suitable for existing and still practised technologies in countries which are in development processes. These HTR specific features offer special possibilities in HTR project financing. Various models are discussed in respect of the special HTR situation. Even if it is not possible to point out in a general manner the best solution - due to national, local and time dependant situations - this paper discusses the HTR specific impacts to buyer's credit financing, supplier's credit financing, barter trades or joint ventures and combined financing. (author). 4 refs, 9 figs

  2. The containment safety of the Dragon Reactor

    International Nuclear Information System (INIS)

    Cullington, G.R.

    1967-08-01

    The original design of the Dragon Reactor was based upon the assumption that fission product emitting fuel elements would be used, leading to two significant considerations. First, a highly active primary circuit would result in normal operation, and second, under accident conditions involving massive core damage and corrosion following a major pressure vessel failure, the bulk of the core burden of fission products would be released. The adoption of coated particle fuel able to retain fission products has changed significantly the philosophy behind the design of the containment. The new philosophy is described and its effect on operating principles is discussed. (UK)

  3. Method for producing dustless graphite spheres from waste graphite fines

    Science.gov (United States)

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  4. Graphite targets at LAMPF

    International Nuclear Information System (INIS)

    Brown, R.D.; Grisham, D.L.

    1983-01-01

    Rotating polycrystalline and stationary pyrolytic graphite target designs for the LAMPF experimental area are described. Examples of finite element calculations of temperatures and stresses are presented. Some results of a metallographic investigation of irradiated pyrolytic graphite target plates are included, together with a brief description of high temperature bearings for the rotating targets

  5. Future Development of Modular HTGR in China after HTR-PM

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Wang, Haitao; Dong Yujie; Li Fu

    2014-01-01

    The modular high temperature gas-cooled reactor (MHTGR) is an inherently safe nuclear energy technology for efficient electricity generation and process heat applications. The MHTGR is promising in China as it may replace fossil fuels in broader energy markets. In line with China’s long-term development plan of nuclear power, the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University developed and designed a MHTGR demonstration plant, named high-temperature gas-cooled reactor-pebble bed module (HTR-PM). The HTR-PM came into the construction phase at the end of 2012. The HTR-PM aims to demonstrate safety, economic potential and modularization technologies towards future commercial applications. Based on experiences obtained from the HTR-PM project with respect to design, manufacture, construction, licensing and project management, a further step aiming to promote commercialization and market applications of the MHTGR is expected. To this purpose, INET is developing a commercialized MHTGR named HTR-PM600 and a conceptual design is under way accordingly. HTR-PM600 is a pebble-bed MHTGR power generation unit with a six-pack of 250MWth reactor modules. The objective is to cogenerate electricity and process heat flexibly and economically in order to meet a variety of market needs. The design of HTR-PM600 closely follows HTR-PM with respect to safety features, system configuration and plant layout. HTR-PM600 has the six modules feeding one steam turbine to generate electricity with capacity to extract high temperature steam from various interfaces of the turbine for further process heat applications. A standard plant consists of two HTR-PM600 units. Based on the economic information of HTR-PM, a preliminary study is carried out on the economic prospect of HTR-PM600. (author)

  6. The verification of DRAGON: progress and lessons learned

    International Nuclear Information System (INIS)

    Marleau, G.

    2002-01-01

    The general requirements for the verification of the legacy code DRAGON are somewhat different from those used for new codes. For example, the absence of a design manual for DRAGON makes it difficult to confirm that the each part of the code performs as required since these requirements are not explicitly spelled out for most of the DRAGON modules. In fact, this conformance of the code can only be assessed, in most cases, by making sure that the contents of the DRAGON data structures, which correspond to the output generated by a module of the code, contains the adequate information. It is also possible in some cases to use the self-verification options in DRAGON to perform additional verification or to evaluate, using an independent software, the performance of specific functions in the code. Here, we will describe the global verification process that was considered in order to bring DRAGON to an industry standard tool-set (IST) status. We will also discuss some of the lessons we learned in performing this verification and present some of the modification to DRAGON that were implemented as a consequence of this verification. (author)

  7. Electrochemical treatment of graphite

    Energy Technology Data Exchange (ETDEWEB)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electrochemical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment, ECT of graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones with respect to the treatment rate and purity (ronghness) of the surface. A small quantity of sludge (6-8%) under ECT is in highly alkali electrolytes.

  8. Electrochemical treatment of graphite

    International Nuclear Information System (INIS)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electroche-- mical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment ECT graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones this is treatment rate and purity (ronghness) of the surface. A sMall quantity of sludge (6-8%) under ECT is in highly alkali electrolytes

  9. The chlamydial periplasmic stress response serine protease cHtrA is secreted into host cell cytosol

    Directory of Open Access Journals (Sweden)

    Flores Rhonda

    2011-04-01

    Full Text Available Abstract Background The periplasmic High Temperature Requirement protein A (HtrA plays important roles in bacterial protein folding and stress responses. However, the role of chlamydial HtrA (cHtrA in chlamydial pathogenesis is not clear. Results The cHtrA was detected both inside and outside the chlamydial inclusions. The detection was specific since both polyclonal and monoclonal anti-cHtrA antibodies revealed similar intracellular labeling patterns that were only removed by absorption with cHtrA but not control fusion proteins. In a Western blot assay, the anti-cHtrA antibodies detected the endogenous cHtrA in Chlamydia-infected cells without cross-reacting with any other chlamydial or host cell antigens. Fractionation of the infected cells revealed cHtrA in the host cell cytosol fraction. The periplasmic cHtrA protein appeared to be actively secreted into host cell cytosol since no other chlamydial periplasmic proteins were detected in the host cell cytoplasm. Most chlamydial species secreted cHtrA into host cell cytosol and the secretion was not inhibitable by a type III secretion inhibitor. Conclusion Since it is hypothesized that chlamydial organisms possess a proteolysis strategy to manipulate host cell signaling pathways, secretion of the serine protease cHtrA into host cell cytosol suggests that the periplasmic cHtrA may also play an important role in chlamydial interactions with host cells.

  10. Extending the temperature range of the HTR

    International Nuclear Information System (INIS)

    Balcomb, J.D.; Wagner, P.

    1975-01-01

    The operating temperature of the high temperature helium-cooled reactor can be increased in a number of ways in order to provide higher temperature nuclear heat for various industrial processes. Modifications are of two types: 1) decrease in the temperature difference between the maximum coated particle fuel temperature and the mean exit gas temperature, and 2) increased maximum coated particle temperature. Gains in the latter category are limited by fission product diffusion into the gas steam and increases greater than 100 0 K are not forseen. Increases in the former category, however, are readily made and a variety of modifications are proposed as follows: incorporation of coated particles in the fuel matrix; use of a more finely-divided fuel coolant hole geometry to increase heat transfer coefficients and reduce conduction temperature differences; large increases in the fuel matrix graphite thermal conductivity (to about 50 W/m 0 K) to reduce conduction temperature differences; and modifications to the core distribution, both radially and axially. By such means the exit gas temperature can be increased to the range of 1200 0 K to 1600 0 K. (author)

  11. Burn-up measurement in the HTR-module-reactor

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

    The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

  12. The safety characteristics of the HTR 500 reactor plant

    International Nuclear Information System (INIS)

    Wachholz, W.

    1987-01-01

    The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

  13. Calculation of HTR-10 first criticality with MVP

    International Nuclear Information System (INIS)

    Xie Jiachun; Yao Lianying

    2015-01-01

    The first criticality of 10 MW pebble-bed high temperature gas-cooled reactor-test module (HTR-10) was calculated with MVP. According to the characteristics of HTR-10, the Statistical Geometry Model of MVP was employed to describe the random arrangement of coated fuel particles in the fuel pebbles and the random distribution of the fuel and dummy pebbles in the core. Compared with previous results from VSOP and MCNP, the MVP results with JENDL-3.3 library were little more different, but the results with ENDF/B-Ⅵ.8 library were very close. The relative errors were less than 0.7%, compared with the first criticality experimental results. The study shows that MVP could be used in the physics calculations for pebble bed high temperature gas-cooled reactors. (authors)

  14. Stress analysis of HTR-10 steam generator heat exchanging tubes

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

    Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

  15. WORD FORMATION ON DRAGON NEST CHAT LANGUAGE

    Directory of Open Access Journals (Sweden)

    Shavitri Cecillia Harsono

    2016-11-01

    Full Text Available Word formation is creation of new words, which sometimes changes a word’s meaning. Words can be formed from multi word phrases as well. In many cases vocabularies in language are formed from combination of words (Haspelmath 2010: 102. Word formation does not only involve changing physical form of the word itself, but also changing the meaning of said word. There are also instances where the physical form retain its original form while the meaning changes. The phenomenon is called semantic change (Stockwell-Minkova 2001:149. In this thesis the research proposed that the said phenomenon occur in virtual environment, such as in MMORPG. Multiplayer online games that feature fantasy setting virtual environment. For the purpose of this research, Dragon Nest South East Asia server was chosen as data source. The samples are taken from players perusing [World] communication channel. The result of the data analysis has shown that the phenomenon of word formation could occur in a virtual environment of MMORPG, specifcally in Dragon Nest SEA. There are two word formation processes found: processes that involve physical changes and processes that do not involve physical changes but rather innate meaning. It is done by both processing daily language vocabulary both physically and changing its innate meaning to create new words that suits the said virtual environment context. This fnding may influence future research on a fresh perspective and untilled feld.

  16. Turbo-machine deployment of HTR-10 GT

    International Nuclear Information System (INIS)

    Zhu Shutang; Wang Jie; Zhang Zhengming; Yu Suyuan

    2005-01-01

    As a testing project of gas turbine modular High Temperature Gas-cooled Reactor (HTGR), HTR-10GT has been studied and developed by Institute of Nuclear and New Energy Technology (INET) of Tsinghua University after the success of HTR-10 with steam turbine cycle. The main purposes of this project are to demonstrate the gas turbine modular HTGR, to optimize the deployment of Power Conversion Unit (PCU) and to verify the techniques of turbo-machine, operating modes and controlling measures. HTR-10GT is concentrated on the PCU design and the turbo-machine deployment. Possible turbo-machine deployments have been investigated and two of them are introduced in this paper. The preliminary design for the turbo-machine of HTR-10GT is single-shaft of vertical layout, arranged by the side of the reactor and the turbo-compressor rotary speed was selected to be 250 s -1 (15000 r/min) by considering the efficiency of turbo-compressor blade systems, the strength conditions and the mass and size characteristics of the turbo-compressor. The rotor system will be supported by electromagnetic bearings (EMBs) to curb the possible pollutions of the primary loop. Of all the components in this design, the high speed turbo-generator seems to be a world-wide technical nut. As an alternative design, a gearbox complex is used to reduce the rotary speed from the turbo-compressor 250 s -1 to 50 s -1 so that the ordinary generator can be used. (authors)

  17. Progress and problems in modelling HTR core dynamics

    International Nuclear Information System (INIS)

    Scherer, W.; Gerwin, H.

    1991-01-01

    In recent years greater effort has been made to establish theoretical models for HTR core dynamics. At KFA Juelich the TINTE (TIme dependent Neutronics and TEmperatures) code system has been developed, which is able to model the primary circuit of an HTR plant using modern numerical techniques and taking into account the mutual interference of the relevant physical variables. The HTR core is treated in 2-D R-Z geometry for both nucleonics and thermo-fluid-dynamics. 2-energy-group diffusion theory is used in the nuclear part including 6 groups of delayed neutron precursors and 14 groups of decay heat producers. Local and non-local heat sources are incorporated, thus simulating gamma ray transport. The thermo-fluid-dynamics module accounts for heterogeneity effects due to the pebble bed structure. Pipes and other components of the primary loop are modelled in 1-D geometry. Forced convection may be treated as well as natural convection in case of blower breakdown accidents. Validation of TINTE has started using the results of a comprehensive experimental program that has been performed at the Arbeitsgemeinschaft Versuchsreaktor GmbH (AVR) high temperature pebble bed reactor at Juelich. In the frame of this program power transients were initiated by varying the helium blower rotational speed or by moving the control rods. In most cases a good accordance between experiment and calculation was found. Problems in modelling the special AVR reactor geometry in two dimensions are described and suggestions for overcoming the uncertainties of experimentally determined control rod reactivities are given. The influence of different polynomial expansions of xenon cross sections to long term transients is discussed together with effects of burnup during that time. Up to now the TINTE code has proven its general applicability to operational core transients of HTR. The effects of water ingress on reactivity, fuel element corrosion and cooling gas properties are now being

  18. Intermediate heat exchanger for HTR process heat application

    International Nuclear Information System (INIS)

    Crambes, M.

    1980-01-01

    In the French study on the nuclear gasification of coal, the following options were recommended: Coal hydrogenation, the hydrogen being derived from CH 4 reforming under the effects of HTR heat; the use of an intermediate helium circuit between the nuclear plant and the reforming plant. The purpose of the present paper is to describe the heat exchanger designed to transfer heat from the primary to the intermediate circuit

  19. Mars Sample Return Landed with Red Dragon

    Science.gov (United States)

    Stoker, Carol R.; Lemke, Lawrence G.

    2013-01-01

    A Mars Sample Return (MSR) mission is the highest priority science mission for the next decade as recommended by the recent Decadal Survey of Planetary Science. However, an affordable program to carry this out has not been defined. This paper describes a study that examined use of emerging commercial capabilities to land the sample return elements, with the goal of reducing mission cost. A team at NASA Ames examined the feasibility of the following scenario for MSR: A Falcon Heavy launcher injects a SpaceX Dragon crew capsule and trunk onto a Trans Mars Injection trajectory. The capsule is modified to carry all the hardware needed to return samples collected on Mars including a Mars Ascent Vehicle (MAV), an Earth Return Vehicle (ERV) and Sample Collection and Storage hardware. The Dragon descends to land on the surface of Mars using SuperSonic Retro Propulsion (SSRP) as described by Braun and Manning [IEEEAC paper 0076, 2005]. Samples are acquired and deliverd to the MAV by a prelanded asset, possibly the proposed 2020 rover. After samples are obtained and stored in the ERV, the MAV launches the sample-containing ERV from the surface of Mars. We examined cases where the ERV is delivered to either low Mars orbit (LMO), C3 = 0 (Mars escape), or an intermediate energy state. The ERV then provides the rest of the energy (delta V) required to perform trans-Earth injection (TEI), cruise, and insertion into a Moon-trailing Earth Orbit (MTEO). A later mission, possibly a crewed Dragon launched by a Falcon Heavy (not part of the current study) retrieves the sample container, packages the sample, and performs a controlled Earth re-entry to prevent Mars materials from accidentally contaminating Earth. The key analysis methods used in the study employed a set of parametric mass estimating relationships (MERs) and standard aerospace analysis software codes modified for the MAV class of launch vehicle to determine the range of performance parameters that produced converged

  20. Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

    OpenAIRE

    Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

    2012-01-01

    Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

  1. Operational requirements of spherical HTR fuel elements and their performance

    International Nuclear Information System (INIS)

    Roellig, K.; Theymann, W.

    1985-01-01

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWel eq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  2. Strengths, weaknesses, opportunities and threats for HTR deployment in Europe

    International Nuclear Information System (INIS)

    Bredimas, Alexandre; Kugeler, Kurt; Fütterer, Michael A.

    2014-01-01

    High temperature nuclear reactors are a technology, of which early versions were demonstrated in the 1960s–1980s in Germany (AVR, THTR) and the United States (Peach Bottom, Fort St. Vrain). HTRs were initially designed for high temperature, high efficiency electricity generation but the technology, the market and the targeted applications have evolved since then to address industrial cogeneration and new operational conditions (in particular new safety regulations). This paper intends to analyse the latest status of HTR today, as regards their intrinsic strengths and weaknesses and their external context, whether positive (opportunities) or negative (threats). Different dimensions are covered by the analysis: technology status, results from R and D programmes (especially in Europe), competences and skills, licensing aspects, experience feedback from demonstrator operation (in particular in Germany), economic conditions and other non-technical aspects. Europe has a comprehensive experience in the field of HTR with capabilities in both pebble bed and prismatic designs (R and D, engineering, manufacturing, operation, dismantling, and the full fuel cycle). Europe is also a promising market for HTR as the process heat market is large with good industrial and cogeneration infrastructures. The analysis of the European situation is to a good deal indicative for the global potential of this technology

  3. Objectives for an HTR R and D physics programme

    Energy Technology Data Exchange (ETDEWEB)

    Johnstone, I; Scott, J A

    1973-10-15

    The paper reviews important objectives for an HTR R and D programme and the importance of particular characteristics for safety and reactor performance is discussed. Uncertainties in the physics characteristics influence reactor design through the inclusion of design margins and contingency allowances and may even eliminate certain design variants. The paper discusses quantitatively the relationship between some important uncertainties and reactor design and operating costs and derives targets for the precision with which it should be possible to compute the corresponding physics characteristics. To extrapolate to power reactor conditions, the need for a comprehensive computational scheme validated by an adequate experimental programme is emphasised. The reduction in uncertainty as the theoretical and experimental reactor physics development proceeds in order to meet the desired target uncertainty is illustrated with respect to the UK's programme in support of the low enriched HTR concept. The current situation for this concept is discussed and compared briefly with that for the Th cycle HTR for which somewhat less zero energy experimental data are available. (auth)

  4. Survey of appropriate endothermic processes for association with the HTR

    International Nuclear Information System (INIS)

    Brown, G.; Harrison, G.E.; Gent, C.W.; Plummer, J.

    1975-01-01

    Emphasis is placed on association of the HTR system as a heat source with chemical processes requiring temperatures up to 850 to 900 0 C, corresponding to a reactor coolant temperature of 950 0 C, though processes requiring temperatures up to 1100 0 C and above are reviewed. Particular attention is given to processes for the production of hydrogen-containing gases, including coal/lignite gasification which has been the subject of a recent study. Rising fuel prices make the HTR an attractive proposition if design concepts and materials can be developed to match the requirements. Other appropriate endothermic processes considered are oil processing, including tar sands and shales, and also energy production. Since the full temperature range of the reactor system must be utilised mention is made of low grade heat uses. Even very large chemical works have relatively small energy requirement by nuclear heat standards and adoption of the HTR as a heat source is likely only if it is associated with a large chemical/metallurgical complex or with the processing of a natural resource. (author)

  5. 7th International Topical Meeting on High Temperature Reactor Technology: The modular HTR is advancing towards reality. Papers and Presentations

    International Nuclear Information System (INIS)

    2014-01-01

    HTR2014 aimed at providing an international platform for researchers, engineers and industrial professionals to share their innovative ideas, valuable experience and recent progresses on high temperature gas-cooled reactor (HTR) and its application technologies.

  6. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

  7. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  8. Dragon exploratory system on Hepatitis C Virus (DESHCV)

    KAUST Repository

    Kwofie, Samuel K.; Radovanovic, Aleksandar; Sundararajan, Vijayaraghava Seshadri; Maqungo, Monique; Christoffels, Alan G.; Bajic, Vladimir B.

    2011-01-01

    text-mining is a useful approach for analyzing the increasing corpus of published scientific literature on HCV. We report here the first comprehensive HCV customized biomedical text-mining based online web resource, dragon exploratory system

  9. Evaluation of Hydraulic Response of the Wave Dragon

    DEFF Research Database (Denmark)

    Frigaard, Peter; Kofoed, Jens Peter

    The present study investigates the hydraulic response of the wave energy converter Wave Dragon. This is done by peforming model tests in a wave tank in the Hydraulics & Coastal Engineering Laboratory at Aalborg University. In the model tests a floating scale model (length scale 1:50) of the Wave...... Dragon is subjected to irregular, long crested irregular and short crested sea conditions corresponding to typical situations under which the Wave Dragon will produce power. Furthermore two situations corresponding to extreme storm conditions are tested. The objective of the study is to determine...... the wave induced forces in the moorings and in the junction between the reflectors and the reservoir part, and motions of the Wave Dragon situated in different sea conditions....

  10. Computer simulation of variform fuel assemblies using Dragon code

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun; Yao Dong

    2005-01-01

    The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

  11. Standalone visualization tool for three-dimensional DRAGON geometrical models

    International Nuclear Information System (INIS)

    Lukomski, A.; McIntee, B.; Moule, D.; Nichita, E.

    2008-01-01

    DRAGON is a neutron transport and depletion code able to solve one-, two- and three-dimensional problems. To date DRAGON provides two visualization modules, able to represent respectively two- and three-dimensional geometries. The two-dimensional visualization module generates a postscript file, while the three dimensional visualization module generates a MATLAB M-file with instructions for drawing the tracks in the DRAGON TRACKING data structure, which implicitly provide a representation of the geometry. The current work introduces a new, standalone, tool based on the open-source Visualization Toolkit (VTK) software package which allows the visualization of three-dimensional geometrical models by reading the DRAGON GEOMETRY data structure and generating an axonometric image which can be manipulated interactively by the user. (author)

  12. New computational methods used in the lattice code DRAGON

    International Nuclear Information System (INIS)

    Marleau, G.; Hebert, A.; Roy, R.

    1992-01-01

    The lattice code DRAGON is used to perform transport calculations inside cells and assemblies for multidimensional geometry using the collision probability method, including the interface current and J ± techniques. Typical geometries that can be treated using this code include CANDU 2-dimensional clusters, CANDU 3-dimensional assemblies, pressurized water reactor (PWR) rectangular and hexagonal assemblies. It contains a self-shielding module for the treatment of microscopic cross section libraries and a depletion module for burnup calculations. DRAGON was written in a modular form in such a way as to accept easily new collision probability options and make them readily available to all the modules that require collision probability matrices like the self-shielding module, the flux solution module and the homogenization module. In this paper the authors present an overview of DRAGON and discuss some of the methods that were implemented in DRAGON in order to improve on its performance

  13. A validation of DRAGON based on lattice experiments

    International Nuclear Information System (INIS)

    Marleau, G.

    1996-01-01

    Here we address the validation of DRAGON using the Chalk River Laboratory experimental database which has already been used for the validation of other codes. Because of the large variety of information for different fuel and moderator types compiled on this database, the most basic modules of DRAGON are thoroughly tested. The general behaviour observed with DRAGON is very good. Its main weakness is seen in the self-shielding ,calculation where the correction applied to the inner fuel pin seems to be overevaluated with respect to the outer fuel pins. One question which is left open this paper concerns the need for inserting end-regions in the DRAGON cells when the heterogeneous B, leakage model is used. (author)

  14. Black swans and dragon kings: A unified model

    Science.gov (United States)

    Eliazar, Iddo

    2017-09-01

    The term “black swan” is a metaphor for outlier events whose statistics are characterized by Pareto's Law and by Zipf's Law; namely, statistics governed by power-law tails. The term “dragon king” is a metaphor for a singular outlier event which, in comparison with all other outlier events, is in a league of its own. As an illustrative example consider the wealth of a family that is sampled at random from a medieval society: the nobility constitutes the black-swan category, and the royal family constitutes the dragon-king category. In this paper we present and analyze a dynamical model that generates, universally and jointly, black swans and dragon kings. According to this model, growing from the microscopic scale to the macroscopic scale, black swans and dragon kings emerge together and invariantly with respect to initial conditions.

  15. EC-funded project (HTR-L) for the definition of a European safety approach for HTR's

    International Nuclear Information System (INIS)

    Ehster, S.; Dominguez, M.T.; Coe, I.; Brinkmann, G.; Lensa, W. von; Mheen, W. van der; Alessandroni, C.; Pirson, J.

    2002-01-01

    The inherent safety features of the HTRs make events leading to severe core damage highly unlikely and constitute the main differentiating aspects compared to LWRs. While a known and stable regulatory environment has long been established for Light Water Reactors, a different approach is necessary for the licensing of HTR based power plants. Among the R and D projects funded by the European Commission for HTR reactors, the HTR-L project is dedicated to the definition of a common and coherent European safety approach and the identification of the main licensing issues for the licensing framework of the Modular HTRs. Other specific objectives of this project are : To develop a methodology to classify the accidental conditions; To define the preliminary requirements for the confinement of radioactive products and to assess the need for a 'conventional' containment structure; To establish a SSC (2) classification and to define the rules for equipment qualification; To identify the key issues that need to be addressed in the licensing process of the HTRs; To organize a workshop with the concerned Safety Authorities at the end of the project. This paper will explain the project objectives and its final expected outcomes. (author)

  16. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  17. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    International Nuclear Information System (INIS)

    Boer, Brian; Ougouag, Abderrafi M.

    2011-01-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no significant

  18. Antibacterial activities of serum from the Komodo Dragon (Varanus komodoensis)

    OpenAIRE

    Mark Merchant; Danyell Henry; Rodolfo Falconi; Bekky Muscher; Judith Bryja

    2013-01-01

    Komodo dragons (Varanus komodoensis) are able to feed on large prey items by injecting a dose of toxic bacteria with their bite that, over time, kills the prey by systemic infection. Dragons also suffer bites from other members of their own species during territorial disputes and feeding frenzies. However, they do not suffer the same fate as their prey, suggesting that they have developed a strong immunity to bacterial infections. This study was undertaken to determine the antibacterial activ...

  19. Wave Dragon Wave Energy Converters Used as Coastal Protection

    DEFF Research Database (Denmark)

    Nørgaard, Jørgen Harck; Andersen, Thomas Lykke; Kofoed, Jens Peter

    2011-01-01

    This paper deals with wave energy converters used to reduce the wave height along shorelines. For this study the Wave Dragon wave energy converter is chosen. The wave height reduction from a single device has been evaluated from physical model tests in scale 1:51.8 of the 260 x 150 m, 24 kW/m model...... Spain, to evaluate the potential for reducing wave heights close the shore by means of Wave Dragons....

  20. Intraerythrocytic iridovirus in central bearded dragons (Pogona vitticeps).

    Science.gov (United States)

    Grosset, Claire; Wellehan, James F X; Owens, Sean D; McGraw, Sabrina; Gaffney, Patricia M; Foley, Janet; Childress, April L; Yun, Susan; Malm, Kirsten; Groff, Joseph M; Paul-Murphy, Joanne; Weber, E Scott

    2014-05-01

    Three adult central bearded dragons (Pogona vitticeps) originating from a commercial breeding facility presented with clinical signs, including anorexia, dehydration, white multifocal lesions on the dorsal aspect of the tongue, blepharospasm, and weight loss. In 1 of 3 lizards, a marked regenerative anemia was noted, and all 3 bearded dragons had erythrocytic intracytoplasmic inclusion bodies. Nine bearded dragons housed in contact also had identical, but fewer intraerythrocytic inclusions. Inclusion bodies examined by electron microscopy had particles consistent with iridoviruses. Attempts to culture the virus were unsuccessful; however, amplification and sequencing of regions of the viral DNA polymerase by polymerase chain reaction confirmed the presence of an iridovirus. One of the bearded dragons died, while the 2 others showing clinical signs were euthanized. The remaining 9 infected bearded dragons of the teaching colony were also euthanized. Postmortem examination revealed a moderate, multifocal, lymphoplasmacytic or mononuclear adenitis of the tongue in the 3 bearded dragons, and a lymphohistiocytic hepatitis with bacterial granulomas in 2 lizards. © 2014 The Author(s).

  1. HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Snoj, Luka [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Lengar, Igor [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Koberl, Oliver [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2014-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  2. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Leland M. Montierth

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  3. In-situ hybridization based quantification of hTR: a possible biomarker in malignant melanoma

    DEFF Research Database (Denmark)

    Vagner, Josephine; Steiniche, Torben; Stougaard, Magnus

    2015-01-01

    thickness suggesting that hTR might be a valuable biomarker in MM. Furthermore, as ISH-based detection requires presence of both hTR and the reverse transcriptase (hTERT) it might be an indicator of active telomerase and thus have future relevance as a predictive biomarker for anti-telomerase treatment....

  4. Evolution of mitochondrial cell death pathway: Proapoptotic role of HtrA2/Omi in Drosophila

    International Nuclear Information System (INIS)

    Igaki, Tatsushi; Suzuki, Yasuyuki; Tokushige, Naoko; Aonuma, Hiroka; Takahashi, Ryosuke; Miura, Masayuki

    2007-01-01

    Despite the essential role of mitochondria in a variety of mammalian cell death processes, the involvement of mitochondrial pathway in Drosophila cell death has remained unclear. To address this, we cloned and characterized DmHtrA2, a Drosophila homolog of a mitochondrial serine protease HtrA2/Omi. We show that DmHtrA2 normally resides in mitochondria and is up-regulated by UV-irradiation. Upon receipt of apoptotic stimuli, DmHtrA2 is translocated to extramitochondrial compartment; however, unlike its mammalian counterpart, the extramitochondrial DmHtrA2 does not diffuse throughout the cytosol but stays near the mitochondria. RNAi-mediated knock-down of DmHtrA2 in larvae or adult flies results in a resistance to stress stimuli. DmHtrA2 specifically cleaves Drosophila inhibitor-of-apoptosis protein 1 (DIAP1), a cellular caspase inhibitor, and induces cell death both in vitro and in vivo as potent as other fly cell death proteins. Our observations suggest that DmHtrA2 promotes cell death through a cleavage of DIAP1 in the vicinity of mitochondria, which may represent a prototype of mitochondrial cell death pathway in evolution

  5. Studi Model Benchmark Mcnp6 Dalam Perhitungan Reaktivitas Batang Kendali Htr-10

    OpenAIRE

    Jupiter S.Pane, Zuhair, Suwoto, Putranto Ilham Yazid

    2016-01-01

    STUDI MODEL BENCHMARK MCNP6 DALAM PERHITUNGAN REAKTIVITAS BATANG KENDALI HTR-10. Dalam operasi reaktor nuklir, sistem batang kendali memainkan peranan yang sangat penting karena didesain untuk mengendalikan reaktivitas teras dan memadamkan reaktor. Nilai reaktivitas batang kendali harus diprediksi secara akurat melalui eksperimen dan perhitungan. Makalah ini mendiskusikan model Benchmark dalam perhitungan reaktivitas batang kendali reaktor HTR-10. Perhitungan dikerjakan dengan program transpo...

  6. Preliminary ripple effect analysis for HTR 350MWt 4 modules construction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, T. H.; Lee, K. Y.; Shin, Y. J. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    We propose quantitative analysis techniques for ripple effects such as the production inducement effect and employment inducement effect for HTR 350MWt x 4 module construction and operation ripple effect based on NOAK. It is known that APR1400 reactors export ripple effect is about 8,500 billion KRW. As a result, HTR construction has more effective effect than that of APR1400.

  7. Vagal innervation is required for pulmonary function phenotype in Htr4-/- mice.

    Science.gov (United States)

    House, John S; Nichols, Cody E; Li, Huiling; Brandenberger, Christina; Virgincar, Rohan S; DeGraff, Laura M; Driehuys, Bastiaan; Zeldin, Darryl C; London, Stephanie J

    2017-04-01

    Human genome-wide association studies have identified over 50 loci associated with pulmonary function and related phenotypes, yet follow-up studies to determine causal genes or variants are rare. Single nucleotide polymorphisms in serotonin receptor 4 ( HTR4 ) are associated with human pulmonary function in genome-wide association studies and follow-up animal work has demonstrated that Htr4 is causally associated with pulmonary function in mice, although the precise mechanisms were not identified. We sought to elucidate the role of neural innervation and pulmonary architecture in the lung phenotype of Htr4 -/- animals. We report here that the Htr4 -/- phenotype in mouse is dependent on vagal innervation to the lung. Both ex vivo tracheal ring reactivity and in vivo flexiVent pulmonary functional analyses demonstrate that vagotomy abrogates the Htr4 -/- airway hyperresponsiveness phenotype. Hyperpolarized 3 He gas magnetic resonance imaging and stereological assessment of wild-type and Htr4 -/- mice reveal no observable differences in lung volume, inflation characteristics, or pulmonary microarchitecture. Finally, control of breathing experiments reveal substantive differences in baseline breathing characteristics between mice with/without functional HTR4 in breathing frequency, relaxation time, flow rate, minute volume, time of inspiration and expiration and breathing pauses. These results suggest that HTR4's role in pulmonary function likely relates to neural innervation and control of breathing. Copyright © 2017 the American Physiological Society.

  8. Toxic leucoencephalopathy after 'chasing the dragon'.

    Science.gov (United States)

    Singh, Rajinder; Saini, Monica

    2015-06-01

    Toxic leucoencephalopathy (TLE) is a rare neurological complication of heroin abuse. 'Chasing the dragon' is an inhalational mode of heroin abuse that originated in Southeast Asia. Intriguingly, no cases of TLE have been reported from this region, although the inhalational mode of heroin abuse is common. We herein report the case of a middle-aged man with a history of polysubstance abuse who presented with progressive neurological symptoms and progressed to an uncommunicative state. While the initial impression was that of iatrogenic parkinsonism, diffuse leucoencephalopathy with sparing of the cerebellum was noted on magnetic resonance imaging. In view of his history of inhalational heroin abuse close to the onset of the neurological symptoms, a diagnosis of TLE was made. No clinical improvement was noted with administration of a dopaminergic agent. This is the first known case of delayed TLE following heroin inhalation from Southeast Asia with the unusual feature of cerebellar sparing.

  9. Carbon-14 Graphitization Chemistry

    Science.gov (United States)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

    Accelerator Mass Spectrometry (AMS) is a process that allows for the analysis of mass of certain materials. It is a powerful process because it results in the ability to separate rare isotopes with very low abundances from a large background, which was previously impossible. Another advantage of AMS is that it only requires very small amounts of material for measurements. An important application of this process is radiocarbon dating because the rare 14C isotopes can be separated from the stable 14N background that is 10 to 13 orders of magnitude larger, and only small amounts of the old and fragile organic samples are necessary for measurement. Our group focuses on this radiocarbon dating through AMS. When performing AMS, the sample needs to be loaded into a cathode at the back of an ion source in order to produce a beam from the material to be analyzed. For carbon samples, the material must first be converted into graphite in order to be loaded into the cathode. My role in the group is to convert the organic substances into graphite. In order to graphitize the samples, a sample is first combusted to form carbon dioxide gas and then purified and reduced into the graphite form. After a couple weeks of research and with the help of various Physics professors, I developed a plan and began to construct the setup necessary to perform the graphitization. Once the apparatus is fully completed, the carbon samples will be graphitized and loaded into the AMS machine for analysis.

  10. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  11. Mechanical design philosophy for the graphite components of the core structure of an HTGR

    International Nuclear Information System (INIS)

    Bodmann, E.

    1987-01-01

    Parallel to the layout and design of the graphite components for THTRs and the succeeding high temperature reactor projects, the design methods for graphite components have been improved over the years. The aim of this works is to develop the design methods which take into account both the particular properties of graphite and the particular functions of the components. Because of the close relation ship between materials and design codes, this development work has progressed with the development, testing and qualification of German reactor graphite. In this paper, the experience in this field of Hochtemperatur Reaktorbau GmbH and the results of the work and approach to the design problems are reported. The example of a HTR 500 design for a 550 MWe power station is taken up, and the core structure is explained. The graphite components are divided into three classes according to the stress limits. The loading of these components is reviewed. The aim of the design is not the complete avoidance of failure, but to avoid the failure of a single component from leading to a disadvantageous consequence which is not allowable. The classification of loading events, Weibull statistics and maximum allowable stress, the formation of the permissible stress, the assessment of stress due to multiaxial loading and so on are described. (Kako, I.)

  12. Development and Reliability Analysis of HTR-PM Reactor Protection System

    International Nuclear Information System (INIS)

    Li Duo; Guo Chao; Xiong Huasheng

    2014-01-01

    High Temperature Gas-Cooled Reactor-Pebble bed Module (HTR-PM) digital Reactor Protection System (RPS) is a dedicated system, which is designed and developed according to HTR-PM NPP protection specifications. To decrease the probability of accident trips and increase the system reliability, HTR-PM RPS has such features as a framework of four redundant channels, two diverse sub-systems in each channel, and two level two-out-of-four logic voters. Reliability analysis of HTR-PM RPS is based on fault tree model. A fault tree is built based on HTR-PM RPS Failure Modes and Effects Analysis (FMEA), and special analysis is focused on the sub-tree of redundant channel ''2-out-of-4'' logic and the fault tree under one channel is bypassed. The qualitative analysis of fault tree, such as RPS weakness according to minimal cut sets, is summarized in the paper. (author)

  13. Serotonergic gene polymorphisms (5-HTTLPR, 5HTR1A, 5HTR2A), and population differences in aggression: traditional (Hadza and Datoga) and industrial (Russians) populations compared.

    Science.gov (United States)

    Butovskaya, Marina L; Butovskaya, Polina R; Vasilyev, Vasiliy A; Sukhodolskaya, Jane M; Fekhredtinova, Dania I; Karelin, Dmitri V; Fedenok, Julia N; Mabulla, Audax Z P; Ryskov, Alexey P; Lazebny, Oleg E

    2018-04-16

    Current knowledge on genetic basis of aggressive behavior is still contradictory. This may be due to the fact that the majority of studies targeting associations between candidate genes and aggression are conducted on industrial societies and mainly dealing with various types of psychopathology and disorders. Because of that, our study was carried on healthy adult individuals of both sex (n = 853). Three populations were examined: two traditional (Hadza and Datoga) and one industrial (Russians), and the association of aggression with the following polymorphisms 5-HTTLPR, rs6295 (5HTR1A gene), and rs6311 (5HTR2A gene) were tested. Aggression was measured as total self-ratings on Buss-Perry Aggression Questionnaire. Distributions of allelic frequencies of 5-HTTLPR and 5HTR1A polymorphisms were significantly different among the three populations. Consequently, the association analyses for these two candidate genes were carried out separately for each population, while for the 5HTR2A polymorphism, it was conducted on the pooled data that made possible to introduce ethnic factor in the ANOVA model. The traditional biometrical approach revealed no sex differences in total aggression in all three samples. The three-way ANOVA (μ + 5-HTTLPR + 5HTR1A + 5HTR2A +ε) with measures of self-reported total aggression as dependent variable revealed significant effect of the second serotonin receptor gene polymorphism for the Hadza sample. For the Datoga, the interaction effect between 5-HTTLPR and 5HTR1A was significant. No significant effects of the used polymorphisms were obtained for Russians. The results of two-way ANOVA with ethnicity and the 5HTR2A polymorphism as main effects and their interactions revealed the highly significant effect of ethnicity, 5HTR2A polymorphism, and their interaction on total aggression. Our data provided obvious confirmation for the necessity to consider the population origin, as well as cultural background of tested individuals, while

  14. Hydraulic Response of the Wave Energy Converter Wave Dragon in Nissum Bredning

    DEFF Research Database (Denmark)

    Kofoed, Jens Peter; Frigaard, Peter

    This report deals with the hydraulic performance of the wave energy converter Wave Dragon, Nissum Bredning prototype.......This report deals with the hydraulic performance of the wave energy converter Wave Dragon, Nissum Bredning prototype....

  15. Nuclear graphite ageing and turnaround

    International Nuclear Information System (INIS)

    Marsden, B.J.; Hall, G.N.; Smart, J.

    2001-01-01

    Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)

  16. Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

    International Nuclear Information System (INIS)

    Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

    2012-01-01

    Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

  17. South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

    International Nuclear Information System (INIS)

    Joubert, J.; Kohtz, N.; Coe, I.

    2008-01-01

    It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /l/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient

  18. Potential of thorium use in the HTR reactor

    International Nuclear Information System (INIS)

    Engelmann, P.; Hansen, U.; Kolb, G.; Leushacke, D.; Teuchert, E.; Werner, H.

    1979-08-01

    In this investigation, several types of reactors and fuel circulations are dealt with as they refer to the region of the Federal Republic of Germany and are compared with each other as to their need for uranium and their costs until 2100. This includes also an investigation covering the effects of a postponed application of uranium-saving reactors, a delayed reprocessing and two variants of the nuclear energy's contribution to electricity generation. After today's light water reactor (LWR) of the pressure water reactor type (DWR) and the sodium-cooled fast breeder (SBR) which is being developed, the technically rather developed helium-cooled high temperature reactor (HTR) is dealt with as another system. The high temperature reactor is, because of its high coolant temperatures, not only suitable as a nuclear power plant, but can also be used to substitute fossile energy sources on the heat market and is being developed in Germany also for use as process heat reactor for nuclear coal gasification. Here the application of nuclear energy is only considered with regard to the region of power generation. Besides the case of the LWR and HTR-operation without reprocessing and fuel recycling for all reactor systems, the calculations also take into consideration the case of the closed fuel recycling. While LWR and SBR are based on the uranium-plutonium-fuel recycling, the thorium-uranium fuel circulation is considered for the HTR with globular fuel elements. As investigations made until today are generally restricted to the system LWR/SBR and the uranium-plutonium circulation, a main concern of the investigations presented here is to show the potential of the Thorium-utilization in high-temperature reactors and to determine how this system can also be applied during the time period concerned to set up a nuclear energy strategy which is safe and profitable as far as the uranium supply is concerned. (orig./UA) 891 UA/orig.- 892 HIS [de

  19. Blood values in wild and captive Komodo dragons (Varanus komodoensis).

    Science.gov (United States)

    Gillespie, Don; Frye, Frederic L.; Stockham, Steven L.; Fredeking, Terry

    2000-01-01

    The Komodo dragon (Varanus komodoensis) is the largest living lizard and occupies a range smaller than that of any other large carnivore in the world. Samples from 33 free-ranging animals at five localities in Komodo National Park, Indonesia were evaluated to assess underlying health problems. To build a comparative database, samples from 44 Komodo dragons in both Indonesian and U.S. zoos were also analyzed. Tests performed included complete blood counts, clinical chemistry profiles, vitamin A, D(3), and E analyses, mineral levels, and screening for chlorinated pesticides or other toxins in wild specimens. Blood samples from wild dragons were positive for hemogregarines, whereas captive specimens were all negative. Total white blood cell counts were consistently higher in captive Komodo dragons than in wild specimens. Reference intervals were established for some chemistry analytes, and values obtained from different groups were compared. Vitamin A and E ranges were established. Vitamin D(3) levels were significantly different in Komodo dragons kept in captive, indoor exhibits versus those with daily ultraviolet-B exposure, whether captive or wild specimens. Corrective measures such as ultraviolet-permeable skylights, direct sunlight exposure, and self-ballasted mercury vapor ultraviolet lamps increased vitamin D(3) concentrations in four dragons to levels comparable with wild specimens. Toxicology results were negative except for background-level chlorinated pesticide residues. The results indicate no notable medical, nutritional, or toxic problems in the wild Komodo dragon population. Problems in captive specimens may relate to, and can be corrected by, husbandry measures such as regular ultraviolet-B exposure. Zoo Biol 19:495-509, 2000. Copyright 2000 Wiley-Liss, Inc.

  20. Does the HTR module have a chance for the future?

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1989-01-01

    The HTR module was developed as a robust and market-orientated heat source for a wide spectrum of applications. Its technology is largely based on that of the AVR. The choice of a low power density and the small core geometry permit thorough use to be made of the favourable safety characteristics and give an extra-ordinarily high degree of passive safety. There are possibilities for its introduction into the international market at present, particularly in the USSR and the People's Republic of China. (orig.)

  1. Recompressed exfoliated graphite articles

    Science.gov (United States)

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2013-08-06

    This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

  2. 77 FR 23125 - Special Local Regulation; Tuscaloosa Dragon Boat Race; Black Warrior River; Tuscaloosa, AL

    Science.gov (United States)

    2012-04-18

    ... commercially transited river system poses significant safety hazards to both the Dragon Boat racers and the...-AA08 Special Local Regulation; Tuscaloosa Dragon Boat Race; Black Warrior River; Tuscaloosa, AL AGENCY... crews, vessels, and persons on navigable waters during the Jr. League of Tuscaloosa Dragon Boat Races...

  3. 78 FR 57063 - Special Local Regulations; Jacksonville Dragon Boat Festival; St. Johns River; Jacksonville, FL

    Science.gov (United States)

    2013-09-17

    ... 1625-AA08 Special Local Regulations; Jacksonville Dragon Boat Festival; St. Johns River; Jacksonville... Jacksonville Dragon Boat Festival, a series of paddle boat races. The event is scheduled to take place on... States during the Jacksonville Dragon Boat Festival. C. Discussion of the Final Rule On Saturday...

  4. Design criteria for high-temperature-affected, metallic and ceramic components, and for the prestressed concrete reactor pressure vessel of future HTR systems. Final report. Vol. 1-4

    International Nuclear Information System (INIS)

    1988-08-01

    This work in five separate volumes reports on the elaboration of basic data for the formulation of design criteria for HTR components and is arranged into the four following subject areas : (1) safety-specific limiting conditions; (2) metallic components; (3) prestressed concrete reactor pressure vessels; (4) graphitic reactor internals. Under item 2, the mechanical and physical characteristics of the materials X20CrMoV 12 1, X10NiCrAlTi 32 20, and NiCr23Co12Mo are examined up to temperatures of 950deg C. Stress-strain rate laws are elaborated for description of the inelastic deformation behavior. The representation of the subject area reactor pressure vessels deals with four main topics: Prestressed concrete support structure, liner, vessel closures, thermal protection system. Quality-assurance classes are defined under item 4 for graphitic components and load levels for load categories. The material evaluation is discussed in detail (e.g. manufacturing monitoring from the raw material to the graphitization and manufacturing testing up to the acceptance test). In addition, the corrosion behavior and irradiation behavior of graphite is examined and rules for computation of stresses in irradiated and unirradiated graphitic components are elaborated. (MM) [de

  5. Bromine intercalated graphite for lightweight composite conductors

    KAUST Repository

    Amassian, Aram; Patole, Archana

    2017-01-01

    A method of fabricating a bromine-graphite/metal composite includes intercalating bromine within layers of graphite via liquid-phase bromination to create brominated-graphite and consolidating the brominated-graphite with a metal nanopowder via a

  6. Graphite and carbon/carbon components for hot gas ducts

    International Nuclear Information System (INIS)

    Popp, G.; Gruber, U.; Boeder, H.; Janssen, K.

    1984-01-01

    The large coal reserves in the Federal Republic of Germany and the uncertainty of the future energy situation on the world market make it appear sound policy to devote some thought to the gasification of coal. For certain chemical processes, moreover, it would be advantageous to have a reasonably priced source of process heat available. In the Federal Republic of Germany this process heat shall be produced in a high-temperature nuclear reactor (HTR), the primary heating temperatures being in the range between 950 deg. C and 1050 deg. C. One serious problem in utilisation of high temperature heat is the resistance of the construction materials. Ceramic materials with high temperature resistance are considered. The material includes graphite and CC carbon fibre reinforced carbon. As a result of the project promoted by Ministerium fur Wirtschaft (Federal Republic of Germany) it has been demonstrated that both CC and graphite manufactured by SIGRI GmbH are well suited for use in high temperature reactors

  7. First Results for Fluid Dynamics, Neutronics and Fission Product Behaviour in HTR applying the HTR Code Package (HCP) Prototype

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Kasselmann, S.; Xhonneux, A.; Lambertz, D.

    2014-01-01

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT are fed back into a new spectrum code of the HCP. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT–3D. Comparisons will be shown against data generated by the legacy codes VSOP99/11, NAKURE and FRESCO-II. (author)

  8. First results for fluid dynamics, neutronics and fission product behavior in HTR applying the HTR code package (HCP) prototype

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J., E-mail: h.j.allelein@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH Aachen University, 52064 Aachen (Germany); Kasselmann, S.; Xhonneux, A.; Tantillo, F.; Trabadela, A.; Lambertz, D. [Forschungszentrum Jülich, 52425 Jülich (Germany)

    2016-09-15

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT lead to new cross sections, which are fed back into MGT-3D. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT-3D. Comparisons will be shown against data generated by SERPENT and the legacy codes VSOP99/11, NAKURE and FRESCO-II.

  9. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  10. Cesium diffusion in graphite

    International Nuclear Information System (INIS)

    Evans, R.B. III; Davis, W. Jr.; Sutton, A.L. Jr.

    1980-05-01

    Experiments on diffusion of 137 Cs in five types of graphite were performed. The document provides a completion of the report that was started and includes a presentation of all of the diffusion data, previously unpublished. Except for data on mass transfer of 137 Cs in the Hawker-Siddeley graphite, analyses of experimental results were initiated but not completed. The mass transfer process of cesium in HS-1-1 graphite at 600 to 1000 0 C in a helium atmosphere is essentially pure diffusion wherein values of (E/epsilon) and ΔE of the equation D/epsilon = (D/epsilon) 0 exp [-ΔE/RT] are about 4 x 10 -2 cm 2 /s and 30 kcal/mole, respectively

  11. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  12. DRAGONS - A Micrometeoroid and Orbital Debris Impact Sensor

    Science.gov (United States)

    Liou, J. -C.; Corsaro, R.; Giovane, F.; Anderson, C.; Sadilek, A.; Burchell, M.; Hamilton, J.

    2015-01-01

    The Debris Resistive/Acoustic Grid Orbital Navy-NASA Sensor (DRAGONS) is intended to be a large area impact sensor for in situ measurements of micrometeoroids and orbital debris (MMOD) in the millimeter or smaller size regime. These MMOD particles are too small to be detected by ground-based radars and optical telescopes, but are still large enough to be a serious safety concern for human space activities and robotic missions in the low Earth orbit (LEO) region. The nominal detection area of a DRAGONS unit is 1 m2, consisting of several independently operated panels. The approach of the DRAGONS design is to combine different particle impact detection principles to maximize information that can be extracted from detected events. After more than 10 years of concept and technology development, a 1 m2 DRAGONS system has been selected for deployment on the International Space Station (ISS) in August 2016. The project team achieved a major milestone when the Preliminary Design Review (PDR) was completed in May 2015. Once deployed on the ISS, this multi-year mission will provide a unique opportunity to demonstrate the MMOD detection capability of the DRAGONS technologies and to collect data to better define the small MMOD environment at the ISS altitude.

  13. SpaceX's Dragon America's next generation spacecraft

    CERN Document Server

    Seedhouse, Erik

    2016-01-01

    This book describes Dragon V2, a futuristic vehicle that not only provides a means for NASA to transport its astronauts to the orbiting outpost but also advances SpaceX’s core objective of reusability. A direct descendant of Dragon, Dragon V2 can be retrieved, refurbished and re-launched. It is a spacecraft with the potential to completely revolutionize the economics of an industry where equipment costing hundreds of millions of dollars is routinely discarded after a single use. It was presented by SpaceX CEO Elon Musk in May 2014 as the spaceship that will carry NASA astronauts to the International Space Station as soon as 2016 SpaceX’s Dragon – America’s Next Generation Spacecraft describes the extraordinary feats of engineering and human achievement that have placed this revolutionary spacecraft at the forefront of the launch industry and positioned it as the precursor for ultimately transporting humans to Mars. It describes the design and development of Dragon, provides mission highlights of the f...

  14. DNA Methylation Analysis of HTR2A Regulatory Region in Leukocytes of Autistic Subjects.

    Science.gov (United States)

    Hranilovic, Dubravka; Blazevic, Sofia; Stefulj, Jasminka; Zill, Peter

    2016-02-01

    Disturbed brain and peripheral serotonin homeostasis is often found in subjects with autism spectrum disorder (ASD). The role of the serotonin receptor 2A (HTR2A) in the regulation of central and peripheral serotonin homeostasis, as well as its altered expression in autistic subjects, have implicated the HTR2A gene as a major candidate for the serotonin disturbance seen in autism. Several studies, yielding so far inconclusive results, have attempted to associate autism with a functional SNP -1438 G/A (rs6311) in the HTR2A promoter region, while possible contribution of epigenetic mechanisms, such as DNA methylation, to HTR2A dysregulation in autism has not yet been investigated. In this study, we compared the mean DNA methylation within the regulatory region of the HTR2A gene between autistic and control subjects. DNA methylation was analysed in peripheral blood leukocytes using bisulfite conversion and sequencing of the HTR2A region containing rs6311 polymorphism. Autistic subjects of rs6311 AG genotype displayed higher mean methylation levels within the analysed region than the corresponding controls (P epigenetic mechanisms might contribute to HTR2A dysregulation observed in individuals with ASD. © 2015 International Society for Autism Research, Wiley Periodicals, Inc.

  15. Genetic variation in HTR4 and lung function: GWAS follow-up in mouse.

    Science.gov (United States)

    House, John S; Li, Huiling; DeGraff, Laura M; Flake, Gordon; Zeldin, Darryl C; London, Stephanie J

    2015-01-01

    Human genome-wide association studies (GWASs) have identified numerous associations between single nucleotide polymorphisms (SNPs) and pulmonary function. Proving that there is a causal relationship between GWAS SNPs, many of which are noncoding and without known functional impact, and these traits has been elusive. Furthermore, noncoding GWAS-identified SNPs may exert trans-regulatory effects rather than impact the proximal gene. Noncoding variants in 5-hydroxytryptamine (serotonin) receptor 4 (HTR4) are associated with pulmonary function in human GWASs. To gain insight into whether this association is causal, we tested whether Htr4-null mice have altered pulmonary function. We found that HTR4-deficient mice have 12% higher baseline lung resistance and also increased methacholine-induced airway hyperresponsiveness (AHR) as measured by lung resistance (27%), tissue resistance (48%), and tissue elastance (30%). Furthermore, Htr4-null mice were more sensitive to serotonin-induced AHR. In models of exposure to bacterial lipopolysaccharide, bleomycin, and allergic airway inflammation induced by house dust mites, pulmonary function and cytokine profiles in Htr4-null mice differed little from their wild-type controls. The findings of altered baseline lung function and increased AHR in Htr4-null mice support a causal relationship between genetic variation in HTR4 and pulmonary function identified in human GWAS. © FASEB.

  16. Intercomparison of graphite irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Hering, H; Perio, P; Seguin, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    While fast neutrons only are effective in damaging graphite, results of irradiations are more or less universally expressed in terms of thermal neutron fluxes. This paper attempts to correlate irradiations made in different reactors, i.e., in fluxes of different spectral compositions. Those attempts are based on comparison of 1) bulk length change and volume expansion, and 2) crystalline properties (e.g., lattice parameter C, magnetic susceptibility, stored energy, etc.). The methods used by various authors for determining the lattice constants of irradiated graphite are discussed. (author)

  17. Survey of HTR related research at IRI, Delft, Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Wallerbos, E.J.M.; Van der Hagen, T.H.J.J.; Van Dam, H. [Interfaculty Reactor Institute IRI, Delft University of Technology, Delft (Netherlands); Tuerkcan, E. [ECN Nuclear Research, Petten (Netherlands)

    1998-09-01

    High temperature helium-cooled reactors have a large potential for inherent safety. Therefore, several projects on HTR research are being carried out or were carried out at the Interfaculty Reactor Institute (IRI) of the Delft University of Technology in Delft, Netherlands. As part of a larger research programme measurements of core reactivity, reactivity worth of safety rods and of small samples being oscillated in the reactor core were carried out at the PROTEUS facility of the Paul Scherrer Institute at Villigen, Switzerland. Together with other partners in the Netherlands a small inherently safe co-generation plant with a pebble-bed HTR core was designed and analysed. It was verified that such a reactor can operate continuously for 10 years by adding continuously fuel pebbles until the maximum available core height is reached. As a new, innovative, inherently safe reactor type the design of a fluidized-bed reactor with coated fuel particles on a helium gas stream is discussed and results are shown for the analysis of inherent criticality safety under varying coolant flow rates. IRI is also taking part in the new IAEA Co-ordinated Research Programme, which involves participation in the start-up experiments of the Japanese HTTR and carrying out calculations for the core physics benchmark test. 11 refs.

  18. HTR System Integration in Europe and South Africa

    International Nuclear Information System (INIS)

    Roelofs, Ferry; Ruer, J.; Cuadrado Garcia, P.; Cetnar, J.; Knoche, D.; Lapins, J.; Kasselman, S.; Stoker, P.; Fütterer, M.

    2014-01-01

    An HTR can be used for production of electricity and process heat. When these two applications are combined, a multitude of systems and components are needed. Whilst meeting the end user needs, this multitude of systems and components has to operate safely and economically. Therefore, within the framework of the European 7th framework program ARCHER project, a design schematic of a nuclear cogeneration system connected to a European and a South African industrial process is established and assessed. In order to provide an objective overview of the different indicators important for decision makers, the main characteristics with respect to the HTR system, the environment, safety, and economics are identified and compared to the characteristics of a modern gas turbine plant. In addition, a gap and SWOT analysis of a nuclear cogeneration system in Europe and South Africa are presented. In order to enable technical analysis of such a nuclear cogeneration system, a multitude of computer codes will be needed. Therefore, a code inventory is established of codes being used in Europe and South Africa for which the requirements for integration, development and qualification are assessed. (author)

  19. Core dynamics of HTR under ATWS and accident conditions

    International Nuclear Information System (INIS)

    Nabbi, R.

    1988-05-01

    The systematic classification of the ATWS has been undertaken by analogy to the considerations made for LWR. The initiating events of ATWS and protection actions of safety systems resulting from monitoring of the system variables have been described. The main emphasis of this work is the analysis of the core dynamic consequences of scram failure during the anticipated transients. The investigation has shown that because of the temperature feedback mechanisms a temperature rise during the ATWS results in a self-shutdown of the reactor. Further inherent safety features of the HTR - conditioned by the high heat capacity of the core and by the compressibility of the coolant - do effectively counteract an undesirable increase of temperature and pressure in the primary circuit. In case of the long-term failure of the forced cooling and following core heatup, neutron physical phenomena appear which determine the reactivity behaviour of the HTR. They are, for instance, the decay of Xenon 135, release of the fission products and subsiding of the top reflector. The results of the computer simulations show that a recriticality has to be excluded during the first 2 days if the reactor is shutdown by the reflector rods at the beginning of the accident. (orig./HP) [de

  20. Market prospects of modular HTR in EEC countries

    International Nuclear Information System (INIS)

    Albisu, F.; Garribba, S.F.; Lefevre, J.C.; Leuchs, D.; Vivante, C.

    1991-01-01

    The energy outlook for the early 21st century is very uncertain. Low-cost oil and natural gas reserves will become seriously depleted and nonfossil energy resources may be urgently required because of environmental reasons. In this framework, small- and medium-size nuclear reactors (SMSNRs), particularly the Modular High-Temperature Reactor (HTR) would allow extension of uses of nuclear energy while being adopted to produce power and/or steam or heat, where heat can be at low or high temperature. For policy making and planning purposes it is meaningful to appraise the market potential of Modular HTR during the next 20 or 30 years. The paper presents the outcomes of country studies on the subject conducted for a sample of EC Member nations, including France, Federal Republic of Germany, Italy, and Spain. Among the goals of the studies are the definition of market segments, and identification of the principal obstacles which will affect future adoption of SMSNRs. Opportunities offered by the different contexts and energy end-uses seem promising. Numerous difficulties and constraints emerge, however, some of which might be eased by actions that national governments or more often the European Community may wish to take. (author)

  1. Market prospects of modular HTR in EEC countries

    International Nuclear Information System (INIS)

    Albisu, F.; Garribba, S.F.; Lefevre, J.C.; Leuchs, D.; Vivante, C.

    1992-01-01

    The energy outlook for the early 21st century is very uncertain. Low-cost oil and natural gas reserves will become seriously depleted and non-fossil energy resources may be urgently required because of environmental reasons. In this framework, the European Economic Community should be able to rely upon nuclear energy as an economic, safe and readily deployable resource for its future. Small and medium-size nuclear reactors (SMSNRs), particularly modular high-temperature reactor (HTR) would allow extension of uses of nuclear energy while being adopted to produce power and/or steam or heat, where heat can be at low or high temperature. For policy making and planning purposes it appears meaningful to appraise the market potential of modular HTR during the next twenty or thirty years. Thus the paper presents the outcomes of country studies on the subject conducted for a sample of Member nations to the European Economic Community including France, Federal Republic of Germany, Italy and Spain. Amongst the goals of the studies are definition of market segments, identification of the principal obstacles which would affect future adoption of SMSNRs. Opportunities offered by the different contexts and energy end-uses seem promising. Numerous difficulties and constraints emerge however, some of which might be eased by actions that national governments or more often the European Economic Community, may wish to take. (orig.)

  2. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  3. The nucleus accumbens 5-HTR4-CART pathway ties anorexia to hyperactivity

    Science.gov (United States)

    Jean, A; Laurent, L; Bockaert, J; Charnay, Y; Dusticier, N; Nieoullon, A; Barrot, M; Neve, R; Compan, V

    2012-01-01

    In mental diseases, the brain does not systematically adjust motor activity to feeding. Probably, the most outlined example is the association between hyperactivity and anorexia in Anorexia nervosa. The neural underpinnings of this ‘paradox', however, are poorly elucidated. Although anorexia and hyperactivity prevail over self-preservation, both symptoms rarely exist independently, suggesting commonalities in neural pathways, most likely in the reward system. We previously discovered an addictive molecular facet of anorexia, involving production, in the nucleus accumbens (NAc), of the same transcripts stimulated in response to cocaine and amphetamine (CART) upon stimulation of the 5-HT4 receptors (5-HTR4) or MDMA (ecstasy). Here, we tested whether this pathway predisposes not only to anorexia but also to hyperactivity. Following food restriction, mice are expected to overeat. However, selecting hyperactive and addiction-related animal models, we observed that mice lacking 5-HTR1B self-imposed food restriction after deprivation and still displayed anorexia and hyperactivity after ecstasy. Decryption of the mechanisms showed a gain-of-function of 5-HTR4 in the absence of 5-HTR1B, associated with CART surplus in the NAc and not in other brain areas. NAc-5-HTR4 overexpression upregulated NAc-CART, provoked anorexia and hyperactivity. NAc-5-HTR4 knockdown or blockade reduced ecstasy-induced hyperactivity. Finally, NAc-CART knockdown suppressed hyperactivity upon stimulation of the NAc-5-HTR4. Additionally, inactivating NAc-5-HTR4 suppressed ecstasy's preference, strengthening the rewarding facet of anorexia. In conclusion, the NAc-5-HTR4/CART pathway establishes a ‘tight-junction' between anorexia and hyperactivity, suggesting the existence of a primary functional unit susceptible to limit overeating associated with resting following homeostasis rules. PMID:23233022

  4. Graphite-based photovoltaic cells

    Science.gov (United States)

    Lagally, Max; Liu, Feng

    2010-12-28

    The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

  5. The present state of the HTR concept based on experience gained from AVR and THTR

    International Nuclear Information System (INIS)

    Wachholz, W.

    1989-01-01

    During the past ten years the development of a specific HTR concept has made remarkable progress. This has been mainly characterized by making use of the safety characteristics typical of the High-Temperature Reactor (HTR). In the design, construction and operation of High-Temperature Reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - comprehensive experience has been gained in the field of operational availability and safety, accident topology and plant risk of HTRs in recent years. This experience is relevant for the entire HTR line independent of specific projects. (author). 3 refs, 5 figs, 1 tab

  6. Simulation of Thermal-hydraulic Process in Reactor of HTR-PM

    International Nuclear Information System (INIS)

    Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

    2014-01-01

    This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)

  7. Applications and Prospects of Modularization Technology in HTR Project Starting from Primary Loop Cavity Construction

    International Nuclear Information System (INIS)

    Yang Guokang; Chen Jing; Huang Wen; Lin Lizhi; Sun Yunlun; Chen Yan; Mao Jiaxin; Wang Yougang; Wang Jinwen; Lin Mingfeng; Yang Mingshan

    2014-01-01

    Primary loop cavity is one of the key areas and major difficulties in HTR-PM project construction. In order to shorten the construction schedule and improve the construction quality, researches on modular design and construction of primary loop cavity has been carried out and the results have been applied in HTR-PM project construction, and got significant application benefit. This paper summarizes the modularization technology application research and project implementation results of primary loop cavity, and analyzes the application and prospects of modularization technology in the HTR project construction. (author)

  8. Design of the steam reformer for the HTR-10 high temperature process heat application

    International Nuclear Information System (INIS)

    Ju Huaiming; Xu Yuanhui; Jia Haijun

    2000-01-01

    The 10 MW High Temperature Reactor Test Module (HTR-10) is being constructed now and planned to be operational in 2000. One of the objectives is to develop the high temperature process heat application. The methane steam reformer is one of the key-facilities for the nuclear process heat application system. The paper describes the conceptual design of the HTR-10 Steam Reformer with He heating, and the design optimization computer code. It can be used to perform sensitivity analysis for parameters, and to improve the design. Principal parameters and construction features of the HTR-10 reformer heated by He are introduced. (author)

  9. MULTIHORMONAL ISLET CELL CARCINOMAS IN THREE KOMODO DRAGONS (VARANUS KOMODOENSIS).

    Science.gov (United States)

    Eustace, Ronan; Garner, Michael M; Cook, Kimberly; Miller, Christine; Kiupel, Matti

    2017-03-01

      Multihormonal pancreatic islet cell carcinomas were found in one female and two male captive geriatric Komodo dragons (Varanus komodoensis). Gross changes in the pancreas were visible in two of the cases. Clinical signs noted in the Komodo dragons were lethargy, weakness, and anorexia. Histologically, the tumors were comprised of nests and cords of well-differentiated neoplastic islet cells with scant amounts of eosinophilic cytoplasm and round, euchromatic nuclei, with rare mitoses. Infiltration by the islet cell tumor into the surrounding acinar tissue was observed in all cases, but no metastatic foci were seen. Multihormone expression was observed in all tumors, which labeled strongly positive for glucagon and somatostatin and focally positive for polypeptide. Pancreatic islet cell neoplasms should be considered in the differential diagnosis for geriatric Komodo dragons presenting with weakness, lethargy, and poor appetite.

  10. Prototype Testing of the Wave Energy Converter Wave Dragon

    DEFF Research Database (Denmark)

    Kofoed, Jens Peter; Frigaard, Peter; Friis-Madsen, Erik

    2006-01-01

    The Wave Dragon is an offshore wave energy converter of the overtopping type. It consists of two wave reflectors focusing the incoming waves towards a ramp, a reservoir for collecting the overtopping water and a number of hydro turbines for converting the pressure head into power. In the period...... from 1998 to 2001 extensive wave tank testing on a scale model was carried at Aalborg University. Then, a 57!27 m wide and 237 tonnes heavy (incl. ballast) prototype of the Wave Dragon, placed in Nissum Bredning, Denmark, was grid connected in May 2003 as the world’s first offshore wave energy...... converter. The prototype is fully equipped with hydro turbines and automatic control systems, and is instrumented in order to monitor power production, wave climate, forces in mooring lines, stresses in the structure and movements of the Wave Dragon. In the period May 2003 to January 2005 an extensive...

  11. Prototype Testing of the Wave Energy Converter Wave Dragon

    DEFF Research Database (Denmark)

    Kofoed, Jens Peter; Frigaard, Peter Bak; Friis-Madsen, Erik

    2004-01-01

    The Wave Dragon is an offshore wave energy converter of the overtopping type. It consists of two wave reflectors focusing the incoming waves towards a ramp, a reservoir for collecting the overtopping water and a number of hydro turbines for converting the pressure head into power. In the period...... from 1998 to 2001 extensive wave tank testing on a scale model was carried at Aalborg University. Then, a 57 x 27 m wide and 237 tonnes heavy (incl. ballast) prototype of the Wave Dragon, placed in Nissum Bredning, Denmark, was grid connected in May 2003 as the world's first offshore wave energy...... converter. The prototype is fully equipped with hydro turbines and automatic control systems, and is instrumented in order to monitor power production, wave climate, forces in mooring lines, stresses in the structure and movements of the Wave Dragon. During the last months, extensive testing has started...

  12. Gastric neuroendocrine carcinomas in bearded dragons (Pogona vitticeps).

    Science.gov (United States)

    Ritter, J M; Garner, M M; Chilton, J A; Jacobson, E R; Kiupel, M

    2009-11-01

    This article describes a newly recognized highly malignant neoplastic entity in young bearded dragons (Pogona vitticeps), gastric neuroendocrine carcinomas, which readily metastasize. Ten bearded dragons with histories of anorexia (8), vomiting (3), hyperglycemia (2), and anemia (3) were included in this study. All animals had neoplastic masses in their stomach, with metastasis to the liver. Microscopically, 6 of these neuroendocrine carcinomas were well-differentiated and 4 were poorly differentiated. For further characterization, immunohistochemistry for protein gene product 9.5, neuron-specific enolase, endorphin, chromogranins A and B, synaptophysin, somatostatin, insulin, glucagon, gastrin, pancreatic polypeptide, and vasoactive intestinal peptide was performed on 5 animals. Because only immunolabeling for somatostatin was consistently observed in all neoplasms, a diagnosis of somatostatinoma was made for these 5 bearded dragons. Some neoplasms also exhibited multihormonal expression. Electron microscopy performed on 1 tumor confirmed the presence of neuroendocrine granules within neoplastic cells. Gastric neuroendocrine carcinomas, and specifically somatostatinomas, have not been previously reported in bearded dragons, or other reptiles, and may be underdiagnosed due to inconsistent, ambiguous clinical signs. In humans, pancreatic somatostatinomas are associated with a syndrome of hypersomatostatinemia, which includes hyperglycemia, weight loss, and anemia, as observed in some of these bearded dragons. Somatostatinomas in humans are commonly associated with neurofibromatosis type 1 (Von Recklinghausen's disease), caused by a mutation in the tumor suppressor gene NF1, which results in decreased expression of neurofibromin. In all 5 animals examined, neoplasms exhibited decreased neurofibromin expression compared with control tissues, suggesting that decreased functional neurofibromin may play a role in the pathogenesis of somatostatinomas in bearded dragons.

  13. Survivability Mode and Extreme Loads on the Mooring Lines of the Wave Dragon Wave Energy Converter

    DEFF Research Database (Denmark)

    Parmeggiani, Stefano; Kofoed, Jens Peter

    This report is a product of the cooperation agreement between Wave Dragon and Aalborg University regarding phase 2 of the development of the Wave Dragon Wave Energy Converter. The research is carried out by testing the 1:51.8 scale model of the Wave Dragon, aiming at the assessment of the surviva......This report is a product of the cooperation agreement between Wave Dragon and Aalborg University regarding phase 2 of the development of the Wave Dragon Wave Energy Converter. The research is carried out by testing the 1:51.8 scale model of the Wave Dragon, aiming at the assessment...... of the department of Civil Engineering at Aalborg University. The outcome of the research will be used as input for future research work aimed at the design of the mooring system and the certification of the structural design for the full scale Wave Dragon demonstrator....

  14. Electronic properties of graphite

    International Nuclear Information System (INIS)

    Schneider, J.

    2010-10-01

    In this thesis, low-temperature magneto-transport (T ∼ 10 mK) and the de Haas-van Alphen effect of both natural graphite and highly oriented pyrolytic graphite (HOPG) are examined. In the first part, low field magneto-transport up to B = 11 T is discussed. A Fourier analysis of the background removed signal shows that the electric transport in graphite is governed by two types of charge carriers, electrons and holes. Their phase and frequency values are in agreement with the predictions of the SWM-model. The SWM-model is confirmed by detailed band structure calculations using the magnetic field Hamiltonian of graphite. The movement of the Fermi at B > 2 T is calculated self-consistently assuming that the sum of the electron and hole concentrations is constant. The second part of the thesis deals with high field magneto-transport of natural graphite in the magnetic field range 0 ≤ B ≤ 28 T. Both spin splitting of magneto-transport features in tilted field configuration and the onset of the charge density wave (CDW) phase for different temperatures with the magnetic field applied normal to the sample plane are discussed. Concerning the Zeeman effect, the SWM calculations including the Fermi energy movement require a g-factor of g* equal to 2.5 ± 0.1 to reproduce the spin spilt features. The measurements of the charge density wave state confirm that its onset magnetic field can be described by a Bardeen-Cooper-Schrieffer (BCS)-type formula. The measurements of the de Haas-van Alphen effect are in agreement with the results of the magneto-transport measurements at low field. (author)

  15. Dragones, serpientes y cocodrilos infernales en la comedia de santos

    OpenAIRE

    Gonzalez Fernandez , Luis

    2009-01-01

    International audience; Se examinan en este artículo algunos casos en los que aparecen dragones, serpientes y cocodrilos vinculados al personaje teatral del demonio en la comedia de santos. El corpus es mayormente calderoniano.; Parcours retraçant dans la comedia de santos (pièces hagiographiques) la présence de monstres tels que les dragons, les serpents et les crocodiles là où il y a une relation spécifique avec le diable. Le corpus examiné porte essentiellement sur les oeuvres de Pedro Cal...

  16. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  17. Cerebral xanthomatosis in three green water dragons (Physignathus cocincinus).

    Science.gov (United States)

    Kummrow, Maya S; Berkvens, Charlene N; Paré, Jean A; Smith, Dale A

    2010-03-01

    Cerebral xanthomatosis was diagnosed in three female green water dragons (Physignathus cocincinus), all of which presented with progressive neurologic signs. No antemortem evidence for xanthomatosis was identified, but on postmortem examination cholesterol granulomas, composed of cholesterol clefts surrounded by macrophages and multinucleated giant cells, were found in the forebrain of each animal and were associated with significant displacement and pressure on the adjacent brain. Although the cause of xanthomatosis in these animals is unknown, nutrition and trauma may be involved in the pathogenesis of this condition. Cerebrum, cholesterol, green water dragon, Physignathus cocincinus, xanthoma.

  18. Research on new BPM used to 'Dragon I'

    International Nuclear Information System (INIS)

    Xu Tiezheng; Xie Yutong; Gao Feng; Dai Wenhua; Gu Zhanjun; Wang Liming; Wang Huacen; Li Jing

    2006-01-01

    The principle of beam position monitoring of button was introduced briefly. It was compared with beam bugs in principle. Based on the result in simulation experiment, a new structure of button was design, and some mistakes in the primary design were corrected. In the really beam experiment of 'Dragon I', the beam waveform and position were monitored. Compared the position curve between button and beam bugs that indicated the data that got from button is credible. In the experiment, the button has an accuracy of 0.5 mm, which is adequate for beam position measurement of 'Dragon I'. (authors)

  19. Dragon-Kings, Black-Swans and Prediction (Invited)

    Science.gov (United States)

    Sornette, D.

    2010-12-01

    Extreme fluctuations or events are often associated with power law statistics. Indeed, it is a popular belief that "wild randomness'' is deeply associated with distributions with power law tails characterized by small exponents. In other words, power law tails are often seen as the epitome of extreme events (the "Black Swan'' story). Here, we document in very different systems that there is life beyond power law tails: power laws can be superseded by "dragon-kings'', monster events that occur beyond (or changing) the power law tail. Dragon-kings reveal hidden mechanisms that are only transiently active and that amplify the normal fluctuations (often described by the power laws of the normal regime). The goal of this lecture is to catalyze the interest of the community of geophysicists across all fields of geosciences so that the "invisible gorilla" fallacy may be avoided. Our own research illustrates that new statistics or representation of data are often necessary to identify dragon-kings, with strategies guided by the underlying mechanisms. Paradoxically, the monsters may be ignored or hidden by the use of inappropriate analysis or statistical tools that amount to cut a mamooth in small pieces, so as to lead to the incorrect belief that only mice exist. In order to stimulate further research, we will document and discuss the dragon-king phenomenon on the statistics of financial losses, economic geography, hydrodynamic turbulence, mechanical ruptures, avalanches in complex heterogeneous media, earthquakes, and epileptic seizures. The special status of dragon-kings open a new research program on their predictability, based on the fact that they belong to a different class of their own and express specific mechanisms amplifying the normal dynamics via positive feedbacks. We will present evidence of these claims for the predictions of material rupture, financial crashes and epileptic seizures. As a bonus, a few remarks will be offered at the end on how the dragon

  20. Testing, Analysis and Control of Wave Dragon, Wave Energy Converter

    DEFF Research Database (Denmark)

    Tedd, James

    of the incident waves upon a wave device allows the possibility of accurately tuning the power-take off mechanism (the hydro-turbines for the Wave Dragon) to capture more energy. A digital filter method for performing this prediction in real-time with minimal computational effort is presented. Construction...... of digital filters is well known within signal processing, but their use for this application in Wave Energy is new. The filter must be designed carefully as the frequency components of waves travel at different speeds. Research presented in this thesis has advanced the development of the Wave Dragon device...

  1. Comparative study of random and uniform models for the distribution of TRISO particles in HTR-10 fuel elements

    International Nuclear Information System (INIS)

    Rosales, J.; Perez, J.; Garcia, C.; Munnoz, A.; Lira, C. A. B. O.

    2015-01-01

    TRISO particles are the specific features of HTR-10 and generally HTGR reactors. Their heterogeneity and random arrangement in graphite matrix of these reactors create a significant modeling challenge. In the simulation of spherical fuel elements using MCNPX are usually created repetitive structures using uniform distribution models. The use of these repetitive structures introduces two major approaches: the non-randomness of the TRISO particles inside the pebbles and the intersection of the pebble surface with the TRISO particles. These approaches could affect significantly the multiplicative properties of the core. In order to study the influence of these approaches in the multiplicative properties was estimated the K inf value in one pebble with white boundary conditions using 4 different configurations regarding the distribution of the TRISO particles inside the pebble: uniform hexagonal model, cubic uniform model, cubic uniform without the effect of cutting and a random distribution model. It was studied the impact these models on core scale solving the problem B1, from the Benchmark Problems presented in a Coordinated Research Program of the IAEA. (Author)

  2. The HTR 500 concept based on pratical THTR and AVR experience

    International Nuclear Information System (INIS)

    Wachholz, W.; Weicht, U.

    1988-01-01

    This paper discusses progress during the past ten years in the development of a specific HTR safety concept. This has been mainly characterized by the abandonment of the LWR specific safety principles and making use of the safety characteristics typical of the high-temperature reactor (HTR). In the design, construction and operation of high-temperature reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - experience has been gained in the field of accident topology and plant risk of HTRs in recent years. This experience, based on detailed accident analyses performed by manufacturers and experts, is relevant for the entire HTR line independent of specific projects. The authors focus on the HTR 500, the first commercial high temperature reactor with a pebble bed core. Its design principles and the design of its systems are based on the earlier AVR and THTR projects

  3. Performance limits of coated particle fuel. Part III. Fission product migration in HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nabielek, H.; Hick, H.; Wagner-Loffler, M.; Voice, E. H.

    1974-06-15

    A general introduction and literature survey to the physics and mathematics of fission product migration in HTR fuel is given as well as a review of available experimental results and their evaluation in terms of models and materials data.

  4. Numerical Simulation of Two-branch Hot Gas Mixing at Reactor Outlet of HTR-PM

    International Nuclear Information System (INIS)

    Hao Pengefei; Zhou Yangping; Li Fu; Shi Lei; He Heng

    2014-01-01

    A series of two-branch model experiment has been finished to investigate the thermal mixing efficiency of the HTR-PM reactor outlet. This paper introduces the numerical simulation on the design of thermal mixing structure of HTR-PM and the test facility with Fluent software. The profiles of temperature, pressure and velocity in the mixing structure design and the test facility are discussed by comparing with the model experiment results. The numerical simulation results of the test facility have good agreement to the experiment results. In addition, the thermal-fluid characters obtained by numerical simulation show the thermal mixing structure of HTR-PM has similarity with the test facility. Finally, it is concluded that the thermal mixing design at HTR-PM reactor outlet can fulfilled the requirements for high thermal mixing efficiency and appropriate pressure drop. (author)

  5. Experiment study on thermal mixing performance of HTR-PM reactor outlet

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2016-09-15

    A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of the test facility is introduced, which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main mixing structure, hot gas duct, exhaust pipe system and I&C system. Experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed, which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

  6. Rework of process effluents from the fabrication of HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lasberg, Ingo; Braehler, Georg [NUKEM Technologies GmbH (Germany); Boyes, David [Pebble Bed Modular Reactor (Pty) Ltd., Centurion (South Africa)

    2008-07-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m{sup 3}/a), isopropanol IPA/water mixtures (130 m{sup 3}/a); Non-Process Water NPW (300 m{sup 3}/a); methanol (7m{sup 3}/a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  7. Rework of process effluents from the fabrication of HTR fuel

    International Nuclear Information System (INIS)

    Lasberg, Ingo; Braehler, Georg; Boyes, David

    2008-01-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m 3 /a), isopropanol IPA/water mixtures (130 m 3 /a); Non-Process Water NPW (300 m 3 /a); methanol (7m 3 /a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  8. HTR-500 - a technical and engineered safeguards concept

    International Nuclear Information System (INIS)

    Schoening, J.; Wachholz, W.; Stoelzl, D.

    1985-01-01

    The plant succeeding the THTR-300 nuclear power plant, which has just started its trial phase of power operation, is the HTR-500. On behalf of the Arbeitsgemeinschaft Hochtemperaturreaktor (AHR), the BBC/HRB Group completed a preliminary project study of a nuclear power plant equipped with a high temperature reactor in the 500 MW power range, in which the changed requirements in the nuclear power market are taken into account and electricity generating costs are to be achieved which are competitive with those of a 1230 MW convoy pressurized water reactor of the present design. On this basis, construction documents are to be drafted, and the licensing procedure under the Atomic Energy Act is to be carried out, within a planning phase of roughly four years. (orig.) [de

  9. 'Once through' cycles in the pebble bed HTR

    International Nuclear Information System (INIS)

    Teuchert, E.

    1977-12-01

    In the pebble bed HTR the 'Once Through' cycles achieve a favorable conservation of uranium resources due to their high burnup and due to the relatively low fissile inventory. A detailed study is given for cycles with highly enriched uranium and thorium, 20% enriched uranium and thorium, and for the low (approximately 8%) enriched cycle. The recommended cycle is based on the known THTR fuel element in the Th/U (93%) cycle. The variant with separate Seed elements and Breed elements presents the best pioneer in view of later recycling and thermal breeding. The minimum proliferation risk is achieved in the Th/U (20%) cycle basing on the fuel element type of the AVR, due to the low amount and high denaturization of the disloaded plutonium. (orig.) [de

  10. Different contributions of HtrA protease and chaperone activities to Campylobacter jejuni stress tolerance and physiology

    DEFF Research Database (Denmark)

    Bæk, Kristoffer Torbjørn; Vegge, Christina Skovgaard; Skórko-Glonek, Joanna

    2011-01-01

    activity is sufficient for growth at high temperature or oxidative stress, whereas the HtrA protease activity is only essential at conditions close to the growth limit for C. jejuni. However, the protease activity was required to prevent induction of the cytoplasmic heat-shock response even at optimal......The microaerophilic bacterium Campylobacter jejuni is the most common cause of bacterial food-borne infections in the developed world. Tolerance to environmental stress relies on proteases and chaperones in the cell envelope such as HtrA and SurA. HtrA displays both chaperone and protease activity......, but little is known about how each of these activities contributes to stress tolerance in bacteria. In vitro experiments showed temperature dependent protease and chaperone activities of C. jejuni HtrA. A C. jejuni mutant lacking only the protease activity of HtrA was used to show that the HtrA chaperone...

  11. Factors influencing selection of a HTR for a developing country

    International Nuclear Information System (INIS)

    Karim, C.S.

    1989-01-01

    Consumption of commercial energy and electricity in Bangladesh has to grow rapidly in order to attain socio-economic development of the country. Nuclear power is considered to be an appropriate proposition due to the inadequacy of indigenous primary energy resources. A technical, economic and financial feasibility study of a 300-500 MWe nuclear power plant is underway now. Responses from different suppliers in SMPR range were enumerated jointly by the Consultants and BAEC under the feasibility study. Criteria for selection of technology and the factor influencing the selection of Modular HTR for Bangladesh are described in the paper. Some indicative results of cost economic calculations are included to help form an idea about various limiting conditions, under which a SMPR with the selected technology could become competitive with the other conventional alternatives. Problems in decision making associated with the uncertainties in estimating plant and fuel cycle costs are enumerated. The implications of not having a reference plant vis-a-vis the advantageous safety features are described to show how these aspects can influence the selection of a new technology like HTR for a developing country. Financing is identifiable as the major problem in implementing a nuclear power project in a developing country like Bangladesh. The entire scope of supplies and services may be broken down into components, so that the burden of financing could be shared by more than one exporting country. Some indicative ideas about the packaging of supplies and services are presented in the paper in order to identify different types of financing sources that could be explored for implementation of the project. Some salient features of the effect of joint-venture on the project financing and implementation are described in the paper. (author). 3 refs, 1 fig

  12. Development of digital I&C system in HTR-PM

    International Nuclear Information System (INIS)

    Shi Guilian

    2014-01-01

    Conclusions: HTR-PM DCS has been under execution for 5 years( 2009-2014) . It has taken CTEC 150 man/year so far. With close cooperation with INET, Chinergyand Shanghai Electric, CTEC overcame difficulties, like iterative design, voluminous customization work, new technology, and lacking of drawings. However, the accomplishment of the planned milestones prepared CTEC for the following work in HTR-PM DCS. 1. The 1ST integrated DCS, including safety DCS, non-safety DCS, DEH supplied by Chinese supplier. Rod control system and DEH are integrated in non-safety DCS. Simplified interface, integrated platform, and easy to use and maintenance. 2. CTEC obtained knowledge of 4th generation HTR-PM digital I&C, key design technology, and riched its DCS products by participation in HTRPM. HTR-PM Safety DCS project provided valuable experience for CTEC’s development and application of FIRMSYS, a safety protection control system platform. 3. The qualification solution by customized HTR-PM safety DCS prototype helps simply safety DCS design, V&V, qualification and safety review of the actual system, but results in some problems in system upgrade and maintenance. With the satisfactory application of FIRMSYS in 1000mw PWR and platform qualification , the future HTR-PM safety DCS could be provided based on a qualified safety DCS platform.

  13. Harwell Graphite Calorimeter

    International Nuclear Information System (INIS)

    Linacre, J.K.

    1970-01-01

    The calorimeter is of the steady state temperature difference type. It contains a graphite sample supported axially in a graphite outer jacket, the assembly being contained in a thin stainless steel outer can. The temperature of the jacket and the temperature difference between sample and jacket are measured by chromel-alumel thermocouples. The instrument is calibrated by means of an electric heater of low mass positioned on the axis of the sample. The resistance of the heater is known and both current through the heater and the potential across it may be measured. The instrument is filled with nitrogen at a pressure of one half atmosphere at room temperature. The calorimeter has been designed for prolonged operation at temperatures up to 600°C, and dose rates up to 1 Wg -1 , and instruments have been in use for periods in excess of one year

  14. Biological monitoring of the micro watershed - Canada del Dragon

    International Nuclear Information System (INIS)

    Fernandez, H.; Rossetti, K.; Caceres, T.; Palma, R.; Garcia, P.; De la Rosa, A.; Seoane, I.

    2012-01-01

    The working group under the Uruguay RCA L-5053 project is about La canada del Dragon (located in Santa Lucia basin - Uruguay) evaluation using benthic macro invertebrates. The main problems of the study area were the riverbanks expansion, the erosion and the pesticides impact on the native fish. The group implemented the water quality evaluation through the study of the invertebrates in the basin.

  15. Ecological allometries and niche use dynamics across Komodo dragon ontogeny.

    Science.gov (United States)

    Purwandana, Deni; Ariefiandy, Achmad; Imansyah, M Jeri; Seno, Aganto; Ciofi, Claudio; Letnic, Mike; Jessop, Tim S

    2016-04-01

    Ontogenetic allometries in ecological habits and niche use are key responses by which individuals maximize lifetime fitness. Moreover, such allometries have significant implications for how individuals influence population and community dynamics. Here, we examined how body size variation in Komodo dragons (Varanus komodoensis) influenced ecological allometries in their: (1) prey size preference, (2) daily movement rates, (3) home range area, and (4) subsequent niche use across ontogeny. With increased body mass, Komodo dragons increased prey size with a dramatic switch from small (≤10 kg) to large prey (≥50 kg) in lizards heavier than 20 kg. Rates of foraging movement were described by a non-linear concave down response with lizard increasing hourly movement rates up until ∼20 kg body mass before decreasing daily movement suggesting reduced foraging effort in larger lizards. In contrast, home range area exhibited a sigmoid response with increased body mass. Intrapopulation ecological niche use and overlap were also strongly structured by body size. Thus, ontogenetic allometries suggest Komodo dragon's transition from a highly active foraging mode exploiting small prey through to a less active sit and wait feeding strategy focused on killing large ungulates. Further, our results suggest that as body size increases across ontogeny, the Komodo dragon exhibited marked ontogenetic niche shifts that enabled it to function as an entire vertebrate predator guild by exploiting prey across multiple trophic levels.

  16. Aeroacoustics of the swinging corrugated tube: Voice of the Dragon

    NARCIS (Netherlands)

    Nakiboglu, G.; Rudenko, O.; Hirschberg, Abraham

    2012-01-01

    When one swings a short corrugated pipe segment around one’s head, it produces a musically interesting whistling sound. As a musical toy it is called a “Hummer” and as a musical instrument, the “Voice of the Dragon.” The fluid dynamics aspects of the instrument are addressed, corresponding to the

  17. Lion and dragon: four centuries of Dutch-Vietnamese relations

    NARCIS (Netherlands)

    Kleinen, J.; van der Zwan, B.; Moors, H.; van Zeeland, T.

    2008-01-01

    Dutch-Vietnamese relations go back as far as the beginning of the seventeenth century. For a long time, relations between the Dutch lion and the Vietnamese dragon have been fragile and even violent. Although the relations were not continuously bad, they remained distant rather than warm. Today

  18. Aeroacoustics of the swinging corrugated tube : voice of the dragon

    NARCIS (Netherlands)

    Nakiboglu, G.; Rudenko, O.; Hirschberg, A.

    2012-01-01

    When one swings a short corrugated pipe segment around one’s head, it produces a musically interesting whistling sound. As a musical toy it is called a "Hummer" and as a musical instrument, the "Voice of the Dragon." The fluid dynamics aspects of the instrument are addressed, corresponding to the

  19. Ecological allometries and niche use dynamics across Komodo dragon ontogeny

    Science.gov (United States)

    Purwandana, Deni; Ariefiandy, Achmad; Imansyah, M. Jeri; Seno, Aganto; Ciofi, Claudio; Letnic, Mike; Jessop, Tim S.

    2016-04-01

    Ontogenetic allometries in ecological habits and niche use are key responses by which individuals maximize lifetime fitness. Moreover, such allometries have significant implications for how individuals influence population and community dynamics. Here, we examined how body size variation in Komodo dragons ( Varanus komodoensis) influenced ecological allometries in their: (1) prey size preference, (2) daily movement rates, (3) home range area, and (4) subsequent niche use across ontogeny. With increased body mass, Komodo dragons increased prey size with a dramatic switch from small (≤10 kg) to large prey (≥50 kg) in lizards heavier than 20 kg. Rates of foraging movement were described by a non-linear concave down response with lizard increasing hourly movement rates up until ˜20 kg body mass before decreasing daily movement suggesting reduced foraging effort in larger lizards. In contrast, home range area exhibited a sigmoid response with increased body mass. Intrapopulation ecological niche use and overlap were also strongly structured by body size. Thus, ontogenetic allometries suggest Komodo dragon's transition from a highly active foraging mode exploiting small prey through to a less active sit and wait feeding strategy focused on killing large ungulates. Further, our results suggest that as body size increases across ontogeny, the Komodo dragon exhibited marked ontogenetic niche shifts that enabled it to function as an entire vertebrate predator guild by exploiting prey across multiple trophic levels.

  20. A Literature Unit for "Dragon's Gate" by Laurence Yep.

    Science.gov (United States)

    Thomas-Vallens, Mary

    Intended as a an aid to classroom teachers, this 52-page handbook presents a literature unit based on the children and young people's book, "Dragon's Gate" by Laurence Yep. It begins with sample lesson plans, pre-reading activities, author information, a book summary, vocabulary lists and suggested vocabulary activities. Next, chapters…

  1. Low Cost Mars Sample Return Utilizing Dragon Lander Project

    Science.gov (United States)

    Stoker, Carol R.

    2014-01-01

    We studied a Mars sample return (MSR) mission that lands a SpaceX Dragon Capsule on Mars carrying sample collection hardware (an arm, drill, or small rover) and a spacecraft stack consisting of a Mars Ascent Vehicle (MAV) and Earth Return Vehicle (ERV) that collectively carry the sample container from Mars back to Earth orbit.

  2. Renovation of the Wave Dragon Nissum Bredning Prototype

    DEFF Research Database (Denmark)

    Tedd, James; Kofoed, Jens Peter; Friis-Madsen, Erik

    2006-01-01

    This paper presents developments of the Wave Dragon, a large offshore wave energy converter. A prototype has been tested in a real sea environment for over 20 months. During 2005 the plant has been in harbor for a major overhaul of several of its components. The motivation for the upgrades...

  3. Here Be Dragons: voorgeschiedenis en ontstaan van Adventure Games

    NARCIS (Netherlands)

    Veugen, Connie

    2004-01-01

    The article traces the history of adventure games from the birth of the fantasy genre in William Morris' work and the origins of the Kriegspiel, through Tolkien's fantasy world and Dungeons and Dragons to the early text adventures and the first graphical adventures.

  4. Vulnerable Cyborgs: Learning to Live with our Dragons.

    NARCIS (Netherlands)

    Coeckelbergh, Mark

    2011-01-01

    Transhumanist visions appear to aim at invulnerability. We are invited to fight the dragon of death and disease, to shed our old, human bodies, and to live on as invulnerable minds or cyborgs. This paper argues that even if we managed to enhance humans in one of these ways, we would remain highly

  5. Data Obtained from Prototype Wave Dragon in Nissum Bredning

    DEFF Research Database (Denmark)

    Tedd, James William; Curie, Marie; Kofoed, Jens Peter

    This report is a product of the Project: Sea Testing and Optimisation of Power Production on a Scale 1:4.5 Test Rig of the Offshore Wave Energy Converter Wave Dragon. This report aims to provide access for the project partners to the raw data obtained from the testing period in Nissum Bredning...

  6. Clones, Drones and Dragons: Ongoing Uncertainties around School Leader Development

    Science.gov (United States)

    Walker, Allan

    2015-01-01

    This article examines a number of key issues around successful school leadership and leader development. Three metaphors are used to frame, track and analyse recent research and commentary in the area--these are clones, drones and dragons. Although development mechanisms rarely fall neatly within one category, the metaphors provide a useful way to…

  7. Editorial: aboard the Red Dragon in 2017 | Thurman | Shakespeare ...

    African Journals Online (AJOL)

    Editorial: aboard the Red Dragon in 2017. Chris Thurman. Abstract. No Abstract. Full Text: EMAIL FULL TEXT EMAIL FULL TEXT · DOWNLOAD FULL TEXT DOWNLOAD FULL TEXT · http://dx.doi.org/10.4314/sisa.v29i1.1 · AJOL African Journals Online. HOW TO USE AJOL... for Researchers · for Librarians · for Authors ...

  8. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Butcher, P.

    1976-11-01

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  9. Design of reactor protection systems for HTR plants generating electric power and process heat problems and solutions

    International Nuclear Information System (INIS)

    Craemer, B.; Dahm, H.; Spillekothen, H.G.

    1982-06-01

    The design basis of the reactor protection system (RPS) for HTR plants generating process heat and electric power is briefly described and some particularities of process heat plants are indicated. Some particularly important or exacting technical measuring positions for the RPS of a process heat HTR with 500 MWsub(th) power (PNP 500) are described and current R + D work explained. It is demonstrated that a particularly simple RPS can be realized in an HTR with modular design. (author)

  10. A standard graphite block

    Energy Technology Data Exchange (ETDEWEB)

    Ivkovic, M; Zdravkovic, Z; Sotic, O [Department of Reactor Physics and Dynamics, Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1966-04-15

    A graphite block was calibrated for the thermal neutron flux of the Ra-Be source using indium foils as detectors. Experimental values of the thermal neutron flux along the central vertical axis of the system were corrected for the self-shielding effect and depression of flux in the detector. The experimental values obtained were compared with the values calculated on the basis of solving the conservation neutron equation by the continuous slowing-down theory. In this theoretical calculation of the flux the Ra-Be source was divided into three resonance energy regions. The measurement of the thermal neutron diffusion length in the standard graphite block is described. The measurements were performed in the thermal neutron region of the system. The experimental results were interpreted by the diffusion theory for point thermal neutron source in the finite system. The thermal neutron diffusion length was calculated to be L= 50.9 {+-}3.1 cm for the following graphite characteristics: density = 1.7 g/cm{sup 3}; boron content = 0.1 ppm; absorption cross section = 3.7 mb.

  11. A standard graphite block

    International Nuclear Information System (INIS)

    Ivkovic, M.; Zdravkovic, Z.; Sotic, O.

    1966-04-01

    A graphite block was calibrated for the thermal neutron flux of the Ra-Be source using indium foils as detectors. Experimental values of the thermal neutron flux along the central vertical axis of the system were corrected for the self-shielding effect and depression of flux in the detector. The experimental values obtained were compared with the values calculated on the basis of solving the conservation neutron equation by the continuous slowing-down theory. In this theoretical calculation of the flux the Ra-Be source was divided into three resonance energy regions. The measurement of the thermal neutron diffusion length in the standard graphite block is described. The measurements were performed in the thermal neutron region of the system. The experimental results were interpreted by the diffusion theory for point thermal neutron source in the finite system. The thermal neutron diffusion length was calculated to be L= 50.9 ±3.1 cm for the following graphite characteristics: density = 1.7 g/cm 3 ; boron content = 0.1 ppm; absorption cross section = 3.7 mb

  12. Dragon enhances BMP signaling and increases transepithelial resistance in kidney epithelial cells.

    Science.gov (United States)

    Xia, Yin; Babitt, Jodie L; Bouley, Richard; Zhang, Ying; Da Silva, Nicolas; Chen, Shanzhuo; Zhuang, Zhenjie; Samad, Tarek A; Brenner, Gary J; Anderson, Jennifer L; Hong, Charles C; Schneyer, Alan L; Brown, Dennis; Lin, Herbert Y

    2010-04-01

    The neuronal adhesion protein Dragon acts as a bone morphogenetic protein (BMP) coreceptor that enhances BMP signaling. Given the importance of BMP signaling in nephrogenesis and its putative role in the response to injury in the adult kidney, we studied the localization and function of Dragon in the kidney. We observed that Dragon localized predominantly to the apical surfaces of tubular epithelial cells in the thick ascending limbs, distal convoluted tubules, and collecting ducts of mice. Dragon expression was weak in the proximal tubules and glomeruli. In mouse inner medullary collecting duct (mIMCD3) cells, Dragon generated BMP signals in a ligand-dependent manner, and BMP4 is the predominant endogenous ligand for the Dragon coreceptor. In mIMCD3 cells, BMP4 normally signaled through BMPRII, but Dragon enhanced its signaling through the BMP type II receptor ActRIIA. Dragon and BMP4 increased transepithelial resistance (TER) through the Smad1/5/8 pathway. In epithelial cells isolated from the proximal tubule and intercalated cells of collecting ducts, we observed coexpression of ActRIIA, Dragon, and BMP4 but not BMPRII. Taken together, these results suggest that Dragon may enhance BMP signaling in renal tubular epithelial cells and maintain normal renal physiology.

  13. Structural disorder of graphite and implications for graphite thermometry

    Science.gov (United States)

    Kirilova, Martina; Toy, Virginia; Rooney, Jeremy S.; Giorgetti, Carolina; Gordon, Keith C.; Collettini, Cristiano; Takeshita, Toru

    2018-02-01

    Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25 megapascal (MPa) and aseismic velocities of 1, 10 and 100 µm s-1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  14. Structural disorder of graphite and implications for graphite thermometry

    Directory of Open Access Journals (Sweden)

    M. Kirilova

    2018-02-01

    Full Text Available Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25  megapascal (MPa and aseismic velocities of 1, 10 and 100 µm s−1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  15. Thermoregulatory behavior and orientation preference in bearded dragons.

    Science.gov (United States)

    Black, Ian R G; Tattersall, Glenn J

    2017-10-01

    The regulation of body temperature is a critical function for animals. Although reliant on ambient temperature as a heat source, reptiles, and especially lizards, make use of multiple voluntary and involuntary behaviors to thermoregulate, including postural changes in body orientation, either toward or away from solar sources of heat. This thermal orientation may also result from a thermoregulatory drive to maintain precise control over cranial temperatures or a rostrally-driven sensory bias. The purpose of this work was to examine thermal orientation behavior in adult and neonatal bearded dragons (Pogona vitticeps), to ascertain its prevalence across different life stages within a laboratory situation and its interaction with behavioral thermoregulation. Both adult and neonatal bearded dragons were placed in a thermal gradient and allowed to voluntarily select temperatures for up to 8h to observe the presence and development of a thermoregulatory orientation preference. Both adult and neonatal dragons displayed a non-random orientation, preferring to face toward a heat source while achieving mean thermal preferences of ~ 33-34°C. Specifically, adult dragons were more likely to face a heat source when at cooler ambient temperatures and less likely at warmer temperatures, suggesting that orientation behavior counter-balances local selected temperatures but contributes to their thermoregulatory response. Neonates were also more likely to select cooler temperatures when facing a heat source, but required more experience before this orientation behavior emerged. Combined, these results demonstrate the importance of orientation to behavioral thermoregulation in multiple life stages of bearded dragons. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. Bromine intercalated graphite for lightweight composite conductors

    KAUST Repository

    Amassian, Aram

    2017-07-20

    A method of fabricating a bromine-graphite/metal composite includes intercalating bromine within layers of graphite via liquid-phase bromination to create brominated-graphite and consolidating the brominated-graphite with a metal nanopowder via a mechanical pressing operation to generate a bromine-graphite/metal composite material.

  17. Chemical stabilization of graphite surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Bistrika, Alexander A.; Lerner, Michael M.

    2018-04-03

    Embodiments of a device, or a component of a device, including a stabilized graphite surface, methods of stabilizing graphite surfaces, and uses for the devices or components are disclosed. The device or component includes a surface comprising graphite, and a plurality of haloaryl ions and/or haloalkyl ions bound to at least a portion of the graphite. The ions may be perhaloaryl ions and/or perhaloalkyl ions. In certain embodiments, the ions are perfluorobenzenesulfonate anions. Embodiments of the device or component including stabilized graphite surfaces may maintain a steady-state oxidation or reduction surface current density after being exposed to continuous oxidation conditions for a period of at least 1-100 hours. The device or component is prepared by exposing a graphite-containing surface to an acidic aqueous solution of the ions under oxidizing conditions. The device or component can be exposed in situ to the solution.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  19. Heat exchanger using graphite foam

    Science.gov (United States)

    Campagna, Michael Joseph; Callas, James John

    2012-09-25

    A heat exchanger is disclosed. The heat exchanger may have an inlet configured to receive a first fluid and an outlet configured to discharge the first fluid. The heat exchanger may further have at least one passageway configured to conduct the first fluid from the inlet to the outlet. The at least one passageway may be composed of a graphite foam and a layer of graphite material on the exterior of the graphite foam. The layer of graphite material may form at least a partial barrier between the first fluid and a second fluid external to the at least one passageway.

  20. Characteristic analysis of rotor dynamics and experiments of active magnetic bearing for HTR-10GT

    International Nuclear Information System (INIS)

    Yang Guojun; Xu Yang; Shi Zhengang; Gu Huidong

    2005-01-01

    A 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) at Tsinghua University of China. The helium turbine and generator system of 10 MW high temperature gas-cooled reactor (HTR-10GT) is the second phase for the HTR-10 project. It is to set up a direct helium cycle to replace the current steam cycle. The active magnetic bearing (AMB) instead of ordinary mechanical bearing was chosen to support the rotor in the HTR-10GT. This rotor is vertically mounted to hold the turbine machine, compressors and the power generator together. The rotor's length is 7 m, its weight is about 1500 kg and the rotating speed is 15000 r/min. The structure of the rotor is so complicated that dynamic analysis of the rotor becomes difficult. One of the challenging problems is to exceed natural frequencies with enough stability and safety during reactor start up, power change and shutdown. The dynamic analysis of the rotor is the base for the design of control system. It is important for the rotor to exceed critical speeds. Some kinds of software and methods, such as MSC.Marc, Ansys, and the Transfer Matrix Method, are compared to fully analyze rotor dynamics characteristic in this paper. The modal analysis has been done for the HTR-10GT rotor. MSC.Marc was finally selected to analyze the vibration mode and the natural frequency of the rotor. The effects of AMB stiffness on the critical speeds of the rotor were studied. The design characteristics of the AMB control system for the HTR-10GT were studied and the related experiment to exceed natural frequencies was introduced. The experimental results demonstrate the system functions and validate the control scheme, which will be used in the HTR-10GT project. (authors)

  1. Purification and preparation of graphite oxide from natural graphite

    Energy Technology Data Exchange (ETDEWEB)

    Panatarani, C., E-mail: c.panatarani@phys.unpad.ac.id; Muthahhari, N.; Joni, I. Made [Instrumentation Systems and Functional Material Processing Laboratory, Department of Physics, Faculty of Mathematics and Natural Sciences, Universitas Padjadjaran, Padjadjaran University, Jl. Raya Bandung-Sumedang KM 21, Jatinangor, 45363, Jawa Barat (Indonesia); Rianto, Anton [Grafindo Nusantara Ltd., Belagio Mall Lantai 2, Unit 0 L3-19, Kawasan Mega Kuningan, Kav. B4 No.3, Jakarta Selatan (Indonesia)

    2016-03-11

    Graphite oxide has attracted much interest as a possible route for preparation of natural graphite in the large-scale production and manipulation of graphene as a material with extraordinary electronic properties. Graphite oxide was prepared by modified Hummers method from purified natural graphite sample from West Kalimantan. We demonstrated that natural graphite is well-purified by acid leaching method. The purified graphite was proceed for intercalating process by modifying Hummers method. The modification is on the reaction time and temperature of the intercalation process. The materials used in the intercalating process are H{sub 2}SO{sub 4} and KMNO{sub 4}. The purified natural graphite is analyzed by carbon content based on Loss on Ignition test. The thermo gravimetricanalysis and the Fouriertransform infrared spectroscopy are performed to investigate the oxidation results of the obtained GO which is indicated by the existence of functional groups. In addition, the X-ray diffraction and energy dispersive X-ray spectroscopy are also applied to characterize respectively for the crystal structure and elemental analysis. The results confirmed that natural graphite samples with 68% carbon content was purified into 97.68 % carbon content. While the intercalation process formed a formation of functional groups in the obtained GO. The results show that the temperature and reaction times have improved the efficiency of the oxidation process. It is concluded that these method could be considered as an important route for large-scale production of graphene.

  2. The Oral and Skin Microbiomes of Captive Komodo Dragons Are Significantly Shared with Their Habitat

    OpenAIRE

    Embriette R. Hyde; Jose A. Navas-Molina; Se Jin Song; Jordan G. Kueneman; Gail Ackermann; Cesar Cardona; Gregory Humphrey; Don Boyer; Tom Weaver; Joseph R. Mendelson; Valerie J. McKenzie; Jack A. Gilbert; Rob Knight; Ashley Shade

    2016-01-01

    ABSTRACT Examining the way in which animals, including those in captivity, interact with their environment is extremely important for studying ecological processes and developing sophisticated animal husbandry. Here we use the Komodo dragon (Varanus komodoensis) to quantify the degree of sharing of salivary, skin, and fecal microbiota with their environment in captivity. Both species richness and microbial community composition of most surfaces in the Komodo dragon?s environment are similar t...

  3. Model Testing of Forces in the Reflector Joint and Mooring Forces on Wave Dragon

    DEFF Research Database (Denmark)

    Gilling, Lasse; Kofoed, Jens Peter; Tedd, James

    This report aims to present the results of a test series analysing the forces in the redesigned reflector joint and the forces in the main mooring link. The resluts presented are intended to be used by WD project partners, for the design and construction of the joint on the prototype Wave Dragon...... at Nissum Bredning and for future North Sea scale Wave Dragon. Lengths, forces and other dimentions presented are scaled to the North sea Wave Dragon unless otherwise specified....

  4. Management of UKAEA graphite liabilities

    International Nuclear Information System (INIS)

    Wise, M.

    2001-01-01

    The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Energy Pile O and Graphite Low Energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6,000 tonnes of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale Pile 1 and the Windscale Advanced Gas-cooled Reactor during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe whilst minimising the associated costs and technical risks. These options include (but are not limited to), disposal as Low Level Waste, incineration, or encapsulation and disposal as Intermediate Level Waste. There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of Comfort' have been sought from both the Low Level Waste and the Intermediate Level Waste disposal organisations to support the

  5. Proceedings of the Fifth Seminar of High Temperature Reactor: The Role and Challenge with HTR Opportunity in the Twenty-first Century

    International Nuclear Information System (INIS)

    As-Natio-Lasman; Zaki-Su'ud; Bambang-Sugiono

    2000-11-01

    The Seminar in HTR Reactor has become routine activities held in BATAN since 1994. This Seminar is a continuation of the Seminar on Technology and HTR Application held by Centre for Development of Advanced Reactor System. The theme of the seminar is Role, Challenge, Opportunity of HTR in the Twenty-first Century. Thirteen papers presented in the seminar were collected into proceedings. The aims of the proceedings is to provide information and references on nuclear technology, mainly on HTR technology. (DII)

  6. Graphite in Science and Nuclear Technique

    OpenAIRE

    Zhmurikov, E. I.; Bubnenkov, I. A.; Dremov, V. V.; Samarin, S. I.; Pokrovsky, A. S.; Harkov, D. V.

    2013-01-01

    The monograph is devoted to the application of graphite and graphite composites in science and technology. The structure and electrical properties, the technological aspects of production of high-strength synthetic graphites, the dynamics of the graphite destruction, traditionally used in the nuclear industry are discussed. It is focuses on the characteristics of graphitization and properties of graphite composites based on carbon isotope 13C. The book is based, generally, on the original res...

  7. Modification of structural graphite machining

    International Nuclear Information System (INIS)

    Lavrenev, M.M.

    1979-01-01

    Studied are machining procedures for structural graphites (GMZ, MG, MG-1, PPG) most widely used in industry, of the article mass being about 50 kg. Presented are dependences necessary for the calculation of cross sections of chip suction tappers and duster pipelines in machine shops for structural graphite machining

  8. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  9. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2012-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  10. Design and Experiment of Auxiliary Bearing for Helium Blower of HTR-PM

    International Nuclear Information System (INIS)

    Yang Guojun; Shi Zhengang; Liu Xingnan; Zhao Jingjing

    2014-01-01

    The helium blower is the important equipment for HTR-PM. Active magnetic bearing (AMB) instead of mechanical bearing is selected to support the rotor of the helium blower. However, one implication of AMB is the requirement to provide the auxiliary bearing to mitigate the effects of failures or overload conditions. The auxiliary bearing is used to support the rotor when the AMB fails to work. It must support the dropping rotor and bear the great impact force and friction heat. The design of the auxiliary bearing is one of the challenging problems in the whole system. It is very important for the helium blower with AMB of HTR-PM to make success. The rotor’s length of helium blower of HTR-PM is about 3.3 m, its weight is about 4000 kg and the rotating speed is 4000 r/min. The axial load is 4500kg, and the radial load is 1950kg. The angular contact ball bearing was selected as the auxiliary bearing. The test rig has been finished. It is difficult to analyze the falling course of the rotor. The preliminary analysis of the dropping rotor was done in the special condition. The impact force of auxiliary bearing was computed for the axial and radial load. And the dropping test of the blower rotor for HTR-10 will be introduced also in this paper. Results offer the important theoretical base for the protector design of the helium blower with AMB for HTR-PM. (author)

  11. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  12. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  13. Glass-Graphite Composite Materials

    International Nuclear Information System (INIS)

    Mayzan, M.Z.H.; Lloyd, J.W.; Heath, P.G.; Stennett, M.C.; Hyatt, N.C.; Hand, R.J.

    2016-01-01

    A summary is presented of investigations into the potential of producing glass-composite materials for the immobilisation of graphite or other carbonaceous materials arising from nuclear power generation. The methods are primarily based on the production of base glasses which are subsequently sintered with powdered graphite or simulant TRISO particles. Consideration is also given to the direct preparation of glass-graphite composite materials using microwave technology. Production of dense composite wasteforms with TRISO particles was more successful than with powdered graphite, as wasteforms containing larger amounts of graphite were resistant to densification and the glasses tried did not penetrate the pores under the pressureless conditions used. Based on the results obtained it is concluded that the production of dense glassgraphite composite wasteforms will require the application of pressure. (author)

  14. HTR1B as a risk profile maker in psychiatric disorders: a review through motivation and memory.

    Science.gov (United States)

    Drago, Antonio; Alboni, Silvia; Brunello, Nicoletta; Nicoletta, Brunello; De Ronchi, Diana; Serretti, Alessandro

    2010-01-01

    Serotonin receptor 1B (HTR1B) is involved in the regulation of the serotonin system, playing different roles in specific areas of the brain. We review the characteristics of the gene coding for HTR1B, its product and the functional role of HTR1B in the neural networks involved in motivation and memory; the central role played by HTR1B in these functions is thoroughly depicted and show HTR1B to be a candidate modulator of the mnemonic and motivationally related symptoms in psychiatric illnesses. In order to challenge this assessment, we analyze how and how much the genetic variations located in the gene that codes for HTR1B impacts on the psychiatric phenotypes by reviewing the literature on this topic. We gathered partial evidence arising from genetic association studies, which suggests that HTR1B plays a relevant role in substance-related and obsessive compulsive disorders. On the other hand, no solid evidence for other psychiatric disorders was found. This finding is quite striking because of the heavy impairment of motivation and of mnemonic-related functions (for example, recall bias) that characterize major psychiatric disorders. The possible reasons for the contrast between the prime relevance of HTR1B in regulating memory and motivation and the limited evidence brought by genetic association studies in humans are discussed, and some suggestions for possible future directions are provided.

  15. 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.

  16. A DRAGON-MCNP comparison of void reactivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Marleau, G [Ecole Polytechnique, Montreal, PQ (Canada). Inst. de Genie Nucleaire; Milgram, M S [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs.

  17. Perception of artificial conspecifics by bearded dragons (Pogona vitticeps).

    Science.gov (United States)

    Frohnwieser, Anna; Pike, Thomas W; Murray, John C; Wilkinson, Anna

    2018-01-09

    Artificial animals are increasingly used as conspecific stimuli in animal behavior research. However, researchers often have an incomplete understanding of how the species under study perceives conspecifics, and hence which features needed for a stimulus to be perceived appropriately. To investigate the features to which bearded dragons (Pogona vitticeps) attend, we measured their lateralized eye use when assessing a successive range of stimuli. These ranged through several stages of realism in artificial conspecifics, to see how features such as color, the presence of eyes, body shape and motion influence behavior. We found differences in lateralized eye use depending on the sex of the observing bearded dragon and the artificial conspecific, as well as the artificial conspecific's behavior. Therefore, this approach can inform the design of robotic animals that elicit biologically-meaningful responses in live animals. This article is protected by copyright. All rights reserved.

  18. A DRAGON-MCNP comparison of void reactivity calculations

    International Nuclear Information System (INIS)

    Marleau, G.

    1995-01-01

    The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs

  19. The simulation of resonance photoneutron produced by dragon-I

    International Nuclear Information System (INIS)

    Xiang Yanjun; Ma Jingfang

    2010-01-01

    The temperature measurement using neutron resonance spectroscopy has many advantages such as non-immerging, inside measurement and local temperature distribution measurement, but the deficiency of high intensity pulsed neutron source limits it's application.In order to study the feasibility of Dragon-I as the pulsed neutron source of temperature measurement, the photoneutron characteristic had been simulated by MCNP5, the photoneutron yield is 1.34 x 10 11 per electron pulse, pulse width is 90ns. the yield is as high as 7.47 x 10 12 per electron pulse when 8cm thick U target had been used, which is only one magnitude lower than the yield of spallation source. the moderation of photoneutron had been simulated using some moderator, the results displayed Dragon-I can be a high intensity,narrow pulse neutron source, it's necessary to study further about it's application to temperature measurement using neutron resonance spectroscopy. (authors)

  20. 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

  1. Fission product behaviour in the primary circuit of an HTR

    International Nuclear Information System (INIS)

    Decken, C.B. von der; Iniotakis, N.

    1981-01-01

    The knowledge of fission product behaviour in the primary circuit of a High Temperature Reactor (HTR) is an essential requirement for the estimations of the availability of the reactor plant in normal operation, of the hazards to personnel during inspection and repair and of the potential danger to the environment from severe accidents. On the basis of the theoretical and experimental results obtained at the ''Institute for Reactor Components'' of the KFA Juelich /1/,/2/ the transport- and deposition behaviour of the fission- and activation products in the primary circuit of the PNP-500 reference plant has been investigated thoroughly. Special work had been done to quantify the uncertainties of the investigations and to calculate or estimate the dose rate level at different components of the primary cooling circuit. The contamination and the dose rate level in the inspection gap in the reactor pressure vessel is discussed in detail. For these investigations in particular the surface structure and the composition of the material, the chemical state of the fission products in the cooling gas, the composition of the cooling gas and the influence of dust on the transport- and deposition behaviour of the fission products have been taken into account. The investigations have been limited to the nuclides Ag-110m; Cs-134 and Cs-137

  2. Axial temperatures and fuel management models for a HTR system

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1971-11-12

    In the HTR system, there is a large difference in temperature between different parts of the reactor core. The softer neutron spectrum in the upper colder core regions tends to shift the power productions in the fresh fuel upwards. As uranium 235 depletes and plutonium with its higher cross sections in the lower hot regions is built-up, an axial power flattening takes place. These effects have been studied in detail for a single column in an equilibrium environment. The aim of this paper is to relate these findings to a whole reactor core and to investigate the influence of axial temperatures on the overall performance and in particular, the fuel management scheme chosen for the reference design. A further objective has been to calculate the reactivity requirements for different part load conditions and for various daily and weekly load diagrams. As the xenon cross section changes significantly with temperature these investigations are performed for an equilibrium core with due representation of axial temperature zones.

  3. Burnup measurement study and prototype development in HTR-PM

    International Nuclear Information System (INIS)

    Yan Weihua; Zhang Zhao; Xiao Zhigang; Zhang Liguo

    2014-01-01

    In a pebble-bed core which employs the multi-pass scheme, it is mandatory to determine the burnup of each pebble after the pebble has been extracted from the core in order to determine whether its design burnup has been reached or whether it has to be reinserted into the core again. The burnup of the fuel pebbles can be determined by measuring the activity of 137 Cs with an HPGe detector because of their good correspondence, which is independent of the irradiation history in the core. Based on experiments and Geant4 simulation, the correction factor between the fuel and calibration source was derived by using the efficiency transfer method. By optimizing spectrum analysis algorithm and parameters, the relative standard deviation of the 137 Cs activity can be still controlled below 3.0% despite of the presence of interfering peaks. On the foundation of the simulation and experiment research, a complete solution for burnup measurement system in HTR-PM is provided. (authors)

  4. The behaviour of spherical HTR fuel elements under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

    1985-07-01

    Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO{sub 2}-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable.

  5. Impact of the Improved Resonance Scattering Kernel on HTR Calculations

    International Nuclear Information System (INIS)

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

    2008-01-01

    The importance of an advanced neutron scattering model for heavy isotopes with strong energy dependent cross sections such as the pronounced resonances of U 238 has been discussed in various publications where the full double differential scattering kernel was derived. In this study we quantify the effect of the new scattering model for specific innovative types of High Temperature Reactor (HTR) systems which commonly exhibit a higher degree of heterogeneity and higher fuel temperatures, hence increasing the importance of the secondary neutron energy distribution. In particular the impact on the multiplication factor (k ∞ ) and the Doppler reactivity coefficient is presented in view of the packing factors and operating temperatures. A considerable reduction of k ∞ (up to 600 pcm) and an increased Doppler reactivity (up to 10%) is observed. An increase of up to 2.3% of the Pu 239 inventory can be noticed at 90 MWd/tHM burnup due to enhanced neutron absorption of U 238 . Those effects are more pronounced for design cases in which the neutron flux spectrum is hardened towards the resolved resonance range. (authors)

  6. Nuclear astrophysics at ISAC with DRAGON: Initial studies

    International Nuclear Information System (INIS)

    Olin, Art; Bishop, Shawn; D'Auria, John M.; Lamey, Michael; Liu, Wenjie; Wrede, Chris; Buchmann, Lothar; Chen, Alan; Hunter, Don; Laird, Alison M.; Ottewell, Dave; Rogers, Joel; Chatterjee, Mohan L.; Engel, Sabine; Strieder, Frank; Gigliotti, Dario; Hussein, Ahmed; Greife, Uwe; Jewett, Cybele; Hutcheon, Dave

    2002-01-01

    The new DRAGON recoil separator facility, designed and built to measure directly the rates of radiative proton and alpha capture reactions important for nuclear astrophysics, is now in operation at the TRIUMF-ISAC radioactive beams facility in Vancouver, Canada. Experiments have been conducted for the first time on the 21Na(p,γ)22Mg reaction. The evolution of nova explosions, and particularly their 22Na abundance, depends sensitively on this reaction rate. The radioactive 21Na beam with an intensity of up to 5 x 108 /s was directed onto a windowless hydrogen gas target (3.8 x 1018 H atoms/cm2). Prompt reaction gamma rays were detected using a BGO array and separated reaction products detected using a silicon strip detector at the end of the 20.8 m recoil mass separator. Yield measurements recording simultaneously singles and coincident signals were performed by scanning in energy over the known resonance reported previously in 22Mg at Ecm = 212 keV, and in addition, over a strong resonance observed at Ecm ≅822 keV. Known resonances in the 21Ne(p,γ)22Na, 20Ne(p,γ)21Na, and 24Mg(p,γ)25Al reactions have been used to calibrate the DRAGON. Studies are in progress to further define the performance of the DRAGON facility. Status of the data analysis and results from system performance studies will be presented along with a brief description of the new ISAC and DRAGON facilities

  7. Analytical study of stress and deformation of HTR fuel blocks

    International Nuclear Information System (INIS)

    Tanaka, M.

    1982-01-01

    A two-dimensional finite element computer code named HANS-GR has been developed to predict the mechanical behavior of the graphite fuel blocks with realistic material properties and core environment. When graphite material is exposed to high temperature and fast neutron flux of high density, strains arise due to thermal expansion, irradiation-induced shrinkage and creep. Thus stresses and distortions are induced in the fuel block in which there are spatial variation of these strains. The analytical method used in the program to predcit these induced stresses and distortions by finite element method is discussed. In order to illustrate the versatility of the computer code, numerical results of two example analyses of the multi-hole type fuel elements in the VHTR Reactor are given. Two example analyses presented are those concerning the stresses in fuel blocks with control rod holes and distortions of the fuel blocks at the periphery of the reactor core. It is considered these phenomena should be carefully examined when the multi-hole type fuel elements are applied to VHTR. It is assured that the predicted mechanical behavior of the graphite components is strongly dependent on the material properties used and obtaining the reliable material property is important to make the analytical prediction a reliable one

  8. Powerful elderly characters in video games: Flemeth of Dragon Age

    Directory of Open Access Journals (Sweden)

    Elisabeta Toma

    2015-12-01

    Full Text Available As games are becoming an increasingly popular medium in various demographic and professional strata, scholars are discussing their content and how they shape society. However, despite an increase in gender analysis of video games, little has been written about orienting games towards an elderly audience, or game representations of aging and older persons. Games specifically designed for older persons are focused on improving cognitive functions, starting from the assumption that the elderly are in need of special games in order to repair age-related deficits. This repair-focused design philosophy comes at the expense of pursuing a broader understanding of quality of life and non-programmatic entertainment. Games-for-fun that also explicitly target the elderly as an audience are almost invisible. In this article we turn our attention to a powerful elderly feminine character in an AAA game designed for entertainment without a serious mission, namely Flemeth from Dragon Age. We discuss how the game depicts and models older characters: What repertoire of portraits has Flemeth as an old woman, in the Dragon Age games? How does Flemeth contribute to an enlarged repertoire of portrayals of old women in video games? We conclude that Flemeth’s gender and age displays in Dragon Age do not impoverish her portrayal but, on the contrary, turn her into a powerful and complex character, thus offering a model for game design to represent and invite older players.

  9. Devriesea agamarum causes dermatitis in bearded dragons (Pogona vitticeps).

    Science.gov (United States)

    Hellebuyck, Tom; Martel, An; Chiers, Koen; Haesebrouck, Freddy; Pasmans, Frank

    2009-03-02

    Devriesea agamarum is frequently isolated from dermatitis in lizards, notably from cheilitis in spiny tailed lizards (genus Uromastyx). It was the aim of the present study to assess the role of this bacterium as a causative agent of dermatitis by fulfilling Koch's postulates. First, its association with diseased lizards was demonstrated. The bacterium was isolated from several, mainly desert dwelling squamate species showing symptoms of dermatitis and/or septicaemia. The affected lizards mainly belonged to the family of the Agamidae (genera Pogona, Uromastyx, Agama) and in one case to the Iguanidae (genus Crotaphytus). Secondly, the occurrence of D. agamarum in 66 clinically healthy bearded dragons, 21 clinically healthy Uromastyx species and 40 squamate eggshells was studied. The bacterium was isolated from the oral cavity of 10 bearded dragons but from none of the healthy Uromastyx species. Hence D. agamarum was found to be part of the oral microbiota in Pogona vitticeps. Finally, bearded dragons (P. vitticeps) were experimentally inoculated with D. agamarum by direct application of a bacterial suspension on intact and abraded skin. At the scarified skin of all inoculated lizards, dermatitis was induced from which D. agamarum was re-isolated. In conclusion, D. agamarum is a facultative pathogenic bacterium, able to cause dermatitis in agamid lizards when the integrity of the skin is breached.

  10. Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

    International Nuclear Information System (INIS)

    Sun Yunxue; Shao Jin

    2015-01-01

    Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

  11. Analysis the Response Function of the HTR Ex-core Neutron Detectors in Different Core Status

    International Nuclear Information System (INIS)

    Fan Kai; Li Fu; Zhou Xuhua

    2014-01-01

    Modular high temperature gas cooled reactor HTR-PM demonstration plant, designed by INET, Tsinghua University, is being built in Shidao Bay, Shandong province, China. HTR-PM adopts pebble bed concept. The harmonic synthesis method has been developed to reconstruct the power distributions on HTR-PM. The method based on the assumption that the neutron detector readings are mainly determined by the status of the core through the power distribution, and the response functions changed little when the status of the core changed. To verify the assumption, the influence factors to the ex-core neutron detectors are calculated in this paper, including the control rod position and the temperature of the core. The results shows that when the status of the core changed, the power distribution changed more remarkable than the response function, but the detector readings could change about 5% because of the response function changing. (author)

  12. Simulation and study on reactivity disturbs dynamic character of HTR-10 nuclear power system

    International Nuclear Information System (INIS)

    Huang Xiaojin; Feng Yuankun

    2002-01-01

    In order to not only know 10 MW High Temperature Gas Cooled Reactor (HTR-10) nuclear power system's dynamic character more deeply but also to satisfy requirements of control system's design and analysis, the dynamic model of HTR-10 nuclear power system is established on the basis of dynamic model of HTR-10 nuclear system, which supplies turbine and generate electricity system model. Using this model, system's main variables' dynamic processes are simulated when control rod takes step reactivity disturb. The concussive progresses which is caused by reactivity disturb are analyzed. The results indicate that fuel temperature changing more slowly than nuclear power makes reactivity negative feedback not to restrain power changing, and then power concussive progress comes to being

  13. Coal conversion and the HTR - basic elements of novel power supply concepts

    International Nuclear Information System (INIS)

    Buerger, F.H.

    1985-01-01

    A meeting under this title was held in Dortmund on 16 to 19 September, 1985, jointly by the VGB Technische Vereinigung der Grosskraftwerksbetreiber e.V., Essen, and the Vereinigte Elektrizitaetswerke Westfalen AG (VEW), Dortmund. The meeting was held in two sections: 'Gersteinwerk power plant - the combination unit K and the KUV coal conversion system' and '7th International conference on HTR technology'. Three technologies were discussed that will have a significant role on the future energy market, i.e., the HTR reactor line (first applied in the Hamm-Uentrop THTR reactor), the new generation of coal-fired power plants with combined gas/steam turbines, and the coal gasification technology. All three systems will make more efficient and less-polluting use of domestic coal by using HTR process heat, by converting coal to widen its range of applications, and by providing more efficient combination units for power plants. (orig./UA) [de

  14. Numerical analysis of magnetically suspended rotor in HTR-10 helium circulator being dropped into auxiliary bearings

    International Nuclear Information System (INIS)

    Zhao Jingxiong; Yang Guojun; Li Yue; Yu Suyuan

    2012-01-01

    Active magnetic bearings (AMB) have been selected to support the rotor of primary helium circulator in commercial 10 Mega-Walt High Temperature Gas-cooled Reactor (HTR-10). In an AMB system, the auxiliary bearings are necessary to protect the AMB components in case of losing power. This paper performs the impact simulation of Magnetically Suspended Rotor in HTR-10 Helium Circulator being dropped into the auxiliary bearings using the finite element program ABAQUS. The dynamic response and the strain field of auxiliary bearings are analyzed. The results achieved by the numerical analysis are in agreement with the experiment results. Therefore, the feasibility of the design of auxiliary bearing and the possibility of using the AMB system in the HTR are proved. (authors)

  15. Concept of a HTR modular plant for generation of process heat in a chemical plant

    International Nuclear Information System (INIS)

    1991-07-01

    This final report summarizes the results of a preliminary study on behalf of Buna AG and Leunawerke AG. With regard to the individual situations the study investigated the conditions for modular HTR-2 reactors to cover on-site process heat and electric power demands. HTR-2 reactor erection and operation were analyzed for their economic efficiency compared with fossil-fuel power plants. Considering the prospective product lines, the technical and economic conditions were developed in close cooperation with Buna AG and Leunawerke AG. The study focused on the technical integration of modular HTR reactors into plants with regard to safety concepts, on planning, acceptance and erection concepts which largely exclude uncalculable scheduling and financial risks, and on comparative economic analyses with regard to fossil-fuel power plants. (orig.) [de

  16. Actual characteristics study on HTR-10GT coupling with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Peng Xuechuang; Zhu Shutang; Wang Jie

    2005-01-01

    HTR-10GT is a testing project coupling the reactor HTR-10 with direct gas turbine cycle. Its thermal cycle can be taken as a closed, recuperated and inter-cooled Brayton cycle. The present study is focused on the thermal cycle performance of HTR-10GT under practical conditions of leakage, pressure losses, etc.. Through thermodynamic analysis, the expression of cycle efficiency for actual thermal cycle is derived. By establishing a physical model with friction loss and leakage, a set of governing equation are constructed based on some reasonable assumptions. The results of actual cycle efficiency have been calculated for different leakage amount at different locations while the effects of leakage under different power level have also been calculated and analyzed. (authors)

  17. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    International Nuclear Information System (INIS)

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-01-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes

  18. Design and application of the HTR-100 industrial nuclear power plant

    International Nuclear Information System (INIS)

    Brandes, S.; Kohl, W.

    1988-01-01

    The small HTR-100 high temperature reactor combines the reactor concept of the AVR reactor, which has been proven for 20 years, with the latest component technology of the THTR power plant which has been in operation since 1985. The nuclear heat supply system is conceived so as to be applicable for the generation of electric power, district heat and process steam according to the customer's demand. The HTR-100 reactor has a thermal power of 258 MW and offers steam parameters of 190 bar/530 0 C. To cover a higher power demand HTR-100 reactors can be combined forming a larger power plant. Economic analyses have shown competitiveness with fossil power plants. (orig.)

  19. Numerical Simulation of Wake Effects in the Lee of a Farm of Wave Dragon Wave Energy Converters

    DEFF Research Database (Denmark)

    Beels, C.; Troch, P.; De Visch, K.

    2009-01-01

    . In this paper wake effects in the lee of a single Wave Dragon WEC and multiple Wave Dragon WECs are studied in a time-dependent mild-slope equation model. The Wave Dragon WEC is a floating offshore converter of the overtopping type. The water volume of overtopped waves is first captured in a basin above mean...

  20. 76 FR 61340 - Notice of Decision To Authorize the Importation of Dragon Fruit From Thailand Into the...

    Science.gov (United States)

    2011-10-04

    ... Inspection Service [Docket No. APHIS-2011-0047] Notice of Decision To Authorize the Importation of Dragon... importation into the continental United States of dragon fruit (multiple genera and species) from Thailand... weeds via the importation of dragon fruit from Thailand. DATES: Effective Date: October 4, 2011. FOR...

  1. 78 FR 42733 - Safety Zone; Cleveland Dragon Boat Festival and Head of the Cuyahoga, Cuyahoga River, Cleveland, OH

    Science.gov (United States)

    2013-07-17

    ...-AA00 Safety Zone; Cleveland Dragon Boat Festival and Head of the Cuyahoga, Cuyahoga River, Cleveland... intended to restrict vessels from a portion of the Cuyahoga River during the Dragon Boat Festival and Head... over a decade and the Dragon Boat Festival for the last 7 years. In response to past years' events, the...

  2. Hypervelocity impacts into graphite

    Science.gov (United States)

    Latunde-Dada, S.; Cheesman, C.; Day, D.; Harrison, W.; Price, S.

    2011-03-01

    Studies have been conducted into the characterisation of the behaviour of commercial graphite (brittle) when subjected to hypervelocity impacts by a range of projectiles. The experiments were conducted with a two-stage gas gun capable of launching projectiles of differing density and strength to speeds of about 6kms-1 at right angles into target plates. The damage caused is quantified by measurements of the crater depth and diameters. From the experimental data collected, scaling laws were derived which correlate the crater dimensions to the velocity and the density of the projectile. It was found that for moderate projectile densities the crater dimensions obey the '2/3 power law' which applies to ductile materials.

  3. Hypervelocity impacts into graphite

    International Nuclear Information System (INIS)

    Latunde-Dada, S; Cheesman, C; Day, D; Harrison, W; Price, S

    2011-01-01

    Studies have been conducted into the characterisation of the behaviour of commercial graphite (brittle) when subjected to hypervelocity impacts by a range of projectiles. The experiments were conducted with a two-stage gas gun capable of launching projectiles of differing density and strength to speeds of about 6kms -1 at right angles into target plates. The damage caused is quantified by measurements of the crater depth and diameters. From the experimental data collected, scaling laws were derived which correlate the crater dimensions to the velocity and the density of the projectile. It was found that for moderate projectile densities the crater dimensions obey the '2/3 power law' which applies to ductile materials.

  4. Calculation of the Fission Product Release for the HTR-10 based on its Operation History

    International Nuclear Information System (INIS)

    Xhonneux, A.; Druska, C.; Struth, S.; Allelein, H.-J.

    2014-01-01

    Since the first criticality of the HTR-10 test reactor in 2000, a rather complex operation history was performed. As the HTR-10 is the only pebble bed reactor in operation today delivering experimental data for HTR simulation codes, an attempt was made to simulate the whole reactor operation up to the presence. Special emphasis was put on the fission product release behaviour as it is an important safety aspect of such a reactor. The operation history has to be simulated with respect to the neutronics, fluid mechanics and depletion to get a detailed knowledge about the time-dependent nuclide inventory. In this paper we report about such a simulation with VSOP 99/11 and our new fission product release code STACY. While STACY (Source Term Analysis Code System) so far was able to calculate the fission product release rates in case of an equilibrium core and during transients, it now can also be applied to running-in-phases. This coupling demonstrates a first step towards an HCP Prototype. Based on the published power histogram of the HTR-10 and additional information about the fuel loading and shuffling, a coupled neutronics, fluid dynamics and depletion calculation was performed. Special emphasis was put on the complex fuel-shuffling scheme within both VSOP and STACY. The simulations have shown that the HTR-10 up to now generated about 2580 MWd while reshuffling the core about 2.3 times. Within this paper, STACY results for the equilibrium core will be compared with FRESCO-II results being published by INET. Compared to these release rates, which are based on a few user defined life histories, in this new approach the fission product release rates of Ag-110m, Cs-137, Sr-90 and I-131 have been simulated for about 4000 tracer pebbles with STACY. For the calculation of the HTR-10 operation history time-dependent release rates are being presented as well. (author)

  5. Reactivity control in HTR power plants with respect to passive safety system. Summary

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-12-01

    The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).

  6. Acoustic emission from polycrystalline graphites

    International Nuclear Information System (INIS)

    Ioka, I.; Yoda, S.; Oku, T.; Miyamoto, Y.

    1987-01-01

    Acoustic emission was monitored from polycrystalline graphites with different microstructure (pore size and pore volume) subjected to compressive loading. The graphites used in this study comprised five brands, that is, PGX, ISEM-1, IG-11, IG-15, and ISO-88. A root mean square (RMS) voltage and event counts of acoustic emission for graphites were measured during compressive loading. The acoustic emission was measured using a computed-based data acquisition and analysis system. The graphites were first deformed up to 80 % of the average fracture stress, then unloaded and reloaded again until the fracture occured. During the first loading, the change in RMS voltage for acoustic emission was detected from the initial stage. During the unloading, the RMS voltage became zero level as soon as the applied stress was released and then gradually rose to a peak and declined. The behavior indicated that the reversed plastic deformation occured in graphites. During the second loading, the RMS voltage gently increased until the applied stress exceeded the maximum stress of the first loading; there is no Kaiser effect in the graphites. A bicrystal model could give a reasonable explanation of this results. The empirical equation between the ratio of σ AE to σ f and σ f was obtained. It is considered that the detection of microfracture by the acoustic emission technique is effective in macrofracture prediction of polycrystalline graphites. (author)

  7. Radiolytic graphite oxidation revisited

    International Nuclear Information System (INIS)

    Minshall, P.C.; Sadler, I.A.; Wickham, A.J.

    1996-01-01

    The importance of radiolytic oxidation in graphite-moderated CO 2 -cooled reactors has long been recognised, especially in the Advanced Gas-Cooled Reactors where potential rates are higher because of the higher gas pressure and ratings than the earlier Magnox designs. In all such reactors, the rate of oxidation is partly inhibited by the CO produced in the reaction and, in the AGR, further reduced by the deliberate addition of CH 4 . Significant roles are also played by H 2 and H 2 O. This paper reviews briefly the mechanisms of these processes and the data on which they are based. However, operational experience has demonstrated that these basic principles are unsatisfactory in a number of respects. Gilsocarbon graphites produced by different manufacturers have demonstrated a significant difference in oxidation rate despite a similar specification and apparent equivalence in their pore size and distribution, considered to be the dominant influence on oxidation rate for a given coolant-gas composition. Separately, the inhibiting influence of CH 4 , which for many years had been considered to arise from the formation of a sacrificial deposit on the pore walls, cannot adequately be explained by the actual quantities of such deposits found in monitoring samples which frequently contain far less deposited carbon than do samples from Magnox reactors where the only source of such deposits is the CO. The paper also describes the current status of moderator weight-loss predictions for Magnox and AGR Moderators and the validation of the POGO and DIFFUSE6 codes respectively. 2 refs, 5 figs

  8. The quest for the understanding of Religious Studies: Seeing dragons

    Directory of Open Access Journals (Sweden)

    Jaco Beyers

    2016-03-01

    Full Text Available Religious Studies is concerned with studying religion or the absence thereof. The concept of religion has been discussed, disliked and dissected over centuries. Some have predicted the disappearance of religion, others have predicted the changing of location from the public to the private sphere and some even the re-emergence of religion. In trying to determine the place and relations of Religious Studies an understanding of what religion entails is necessary. It is clear that Religious Studies consists of a multiform subject field and a variety of disciplines with a multiplicity of issues, interests and topics together with a wide variety of approaches and methods. Some scholars have described religion as a �saturated phenomenon� trying to indicate how the diversity of elements described as religious came to shroud the true subject matter. All these hindrances on the road to comprehending religion are like dragons preventing one from completing a (holy! quest. This article does not want to provide new answers to an old debate. In this sense this article is not an attempt at slaying the dragons but identifying them. Three issues (dragons are discussed. How religion, the object of Religious Studies, should be viewed? What methods are employed by Religious Studies and the relatedness of Religious Studies to Theology? In the end the article wants to provide direction on how Religious Studies, as academic discipline, can collaborate with research in Theology.Intradisciplinary and/or interdisciplinary implications: This article discusses the development of the subject of Religious Studies by providing a historic overview of sociological influences on the development. In this sense this article is not an attempt at slaying the dragons but identifying them. Three issues (dragons are discussed: how religion, the object of Religious Studies, should be viewed; what methods are employed by Religious Studies and the relatedness of Religious Studies to

  9. Influence of Polymorphisms in the HTR3A and HTR3B Genes on Experimental Pain and the Effect of the 5-HT3 Antagonist Granisetron.

    Science.gov (United States)

    Louca Jounger, Sofia; Christidis, Nikolaos; Hedenberg-Magnusson, Britt; List, Thomas; Svensson, Peter; Schalling, Martin; Ernberg, Malin

    2016-01-01

    The aim of this study was to investigate experimentally if 5-HT3 single nucleotide polymorphisms (SNP) contribute to pain perception and efficacy of the 5-HT3-antagonist granisetron and sex differences. Sixty healthy participants were genotyped regarding HTR3A (rs1062613) and HTR3B (rs1176744). First, pain was induced by bilateral hypertonic saline injections (HS, 5.5%, 0.2 mL) into the masseter muscles. Thirty min later the masseter muscle on one side was pretreated with 0.5 mL granisetron (1 mg/mL) and on the other side with 0.5 mL placebo (isotonic saline) followed by another HS injection (0.2 mL). Pain intensity, pain duration, pain area and pressure pain thresholds (PPTs) were assessed after each injection. HS evoked moderate pain, with higher intensity in the women (P = 0.023), but had no effect on PPTs. None of the SNPs influenced any pain variable in general, but compared to men, the pain area was larger in women carrying the C/C (HTR3A) (P = 0.015) and pain intensity higher in women with the A/C alleles (HTR3B) (P = 0.019). Pre-treatment with granisetron reduced pain intensity, duration and area to a lesser degree in women (P granisetron. Women carrying the C/T & T/T (HTR3A) genotype had less reduction of pain intensity (P = 0.041) and area (P = 0.005), and women with the C/C genotype (HTR3B) had less reduction of pain intensity (P = 0.030), duration (P = 0.030) and area compared to men (P = 0.017). In conclusion, SNPs did not influence experimental muscle pain or the effect of granisetron on pain variables in general, but there were some sex differences in pain variables that seem to be influenced by genotypes. However, due to the small sample size further research is needed before any firm conclusions can be drawn.

  10. Influence of Polymorphisms in the HTR3A and HTR3B Genes on Experimental Pain and the Effect of the 5-HT3 Antagonist Granisetron.

    Directory of Open Access Journals (Sweden)

    Sofia Louca Jounger

    Full Text Available The aim of this study was to investigate experimentally if 5-HT3 single nucleotide polymorphisms (SNP contribute to pain perception and efficacy of the 5-HT3-antagonist granisetron and sex differences. Sixty healthy participants were genotyped regarding HTR3A (rs1062613 and HTR3B (rs1176744. First, pain was induced by bilateral hypertonic saline injections (HS, 5.5%, 0.2 mL into the masseter muscles. Thirty min later the masseter muscle on one side was pretreated with 0.5 mL granisetron (1 mg/mL and on the other side with 0.5 mL placebo (isotonic saline followed by another HS injection (0.2 mL. Pain intensity, pain duration, pain area and pressure pain thresholds (PPTs were assessed after each injection. HS evoked moderate pain, with higher intensity in the women (P = 0.023, but had no effect on PPTs. None of the SNPs influenced any pain variable in general, but compared to men, the pain area was larger in women carrying the C/C (HTR3A (P = 0.015 and pain intensity higher in women with the A/C alleles (HTR3B (P = 0.019. Pre-treatment with granisetron reduced pain intensity, duration and area to a lesser degree in women (P < 0.05, but the SNPs did not in general influence the efficacy of granisetron. Women carrying the C/T & T/T (HTR3A genotype had less reduction of pain intensity (P = 0.041 and area (P = 0.005, and women with the C/C genotype (HTR3B had less reduction of pain intensity (P = 0.030, duration (P = 0.030 and area compared to men (P = 0.017. In conclusion, SNPs did not influence experimental muscle pain or the effect of granisetron on pain variables in general, but there were some sex differences in pain variables that seem to be influenced by genotypes. However, due to the small sample size further research is needed before any firm conclusions can be drawn.

  11. The failure mechanisms of HTR coated particle fuel and computer code

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

    2010-01-01

    The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

  12. European energy policy and the potential impact of HTR and nuclear cogeneration

    International Nuclear Information System (INIS)

    Fütterer, Michael A.; Carlsson, Johan; Groot, Sander de; Deffrennes, Marc; Bredimas, Alexandre

    2014-01-01

    This paper first provides an update on the current state of play and the potential future role of nuclear energy in Europe. It then describes the EU energy policy tools in the area of nuclear technology. It explains the three-tier strategy of the European nuclear technology platform and its demonstration initiatives, here specifically for nuclear cogeneration and HTR. The paper closes with an outlook on the boundary conditions at which HTR can become attractive for nuclear cogeneration, not only from an energy policy viewpoint but also economically

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

  14. Maw and spent HTR Fuel Element Test storage in Boreholes in rock salt

    International Nuclear Information System (INIS)

    Barnert, E.; Brucher, P.H.; Kroth, K.; Merz, E.; Niephaus, D.

    1986-01-01

    The Budesminister fur Forschung und Technolgie (BMFT, Federal Ministry for Research and Technology) is sponsoring a project at the Kernforschungsanlage Julich (KFA, Juelich Nuclear Research Centre) entitled ''MAW and HTR Fuel Element Test disposal in Boreholes.'' The aim of this project is to develop a technique for the final disposal of (1) dissolver sludge, (2) cladding hulls/structural components and (3) spent HTR fuels elements in salt, and to test this technique in the abandoned Asse salt mine, including safety calculations and safety engineering demonstrations. The project is divided into the sub-projects I ''Disposal/sealing technique'' and II ''Retrievable disposal test.''

  15. The properties of spherical fuel elements and its behavior in the modular HTR

    International Nuclear Information System (INIS)

    Lohnert, G.H.; Ragoss, H.

    1985-01-01

    The reference fuel element for all future HTR applications in the Federal Republic of Germany as developed by NUKEM/HOBEG in the framework of the 'High temperature Fuel-Cycle Project' had to be scrutinised for its compatibility with all the other design principles of the modular HTR, or possibly for restrictions forced upon reactor layout. This reference fuel element can be characterized by the following features: moulded spherical fuel element of 60 mm in diameter with fuel free shell of 5 mm thickness, based on carbon matrix; low enriched uranium (U/Pu fuel cycle); UO 2 fuel kernels; TRISO coating (pyrocarbon and additional SiC layers)

  16. Fission Product Releases from a Core into a Coolant of a Prismatic 350-MWth HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A prismatic 350-MW{sub th} high temperature reactor (HTR) is a means to generate electricity and process heat for hydrogen production. The HTR will be operated for an extended fuel burnup of more than 150 GWd/MTU. Korea Atomic Energy Research Institute (KAERI) is performing a point design for the HTR which is a pre-conceptual design for the analysis and assessment of engineering feasibility of the reactor. In a prismatic HTR, metallic and gaseous fission products (FPs) are produced in the fuel, moved through fuel materials, and released into a primary coolant. The FPs released into the coolant are deposited on the various helium-wetted surfaces in the primary circuit, or they are sorbed on particulate matters in the primary coolant. The deposited or sorbed FPs are released into the environment through the leakage or venting of the primary coolant. It is necessary to rigorously estimate such radioactivity releases into the environment for securing the health and safety of the occupational personnel and the public. This study treats the FP releases from a core into a coolant of a prismatic 350-MW{sub th} HTR. These results can be utilized as input data for the estimation of FP migration from a coolant into the environment. The analysis of fission product release within a prismatic 350-MW{sub th} HTR has been done. It was assumed that the HTR was operated at constant temperature and power for 1500 EFPDs. - The final burnup is 152 GWd/tHM at packing fraction of 25 %, and the final fast fluence is about 8 X 10{sup 21} n/cm{sup 2}, E{sub n} > 0.1 MeV. - The temperatures at the compact center and at the center of a kernel located at the compact center are 884 and 893 .deg. C, respectively, when the packing fraction is 25 % and the coolant temperature is 850 .deg. C. - Xenon is the most radioactive fission product in a coolant of a prismatic HTR when there are broken TRISOs and fuel component contaminated with heavy metals. For metallic fission products, the radioactivity

  17. European energy policy and the potential impact of HTR and nuclear cogeneration

    Energy Technology Data Exchange (ETDEWEB)

    Fütterer, Michael A., E-mail: michael.fuetterer@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755ZG Petten (Netherlands); Carlsson, Johan [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755ZG Petten (Netherlands); Groot, Sander de [Nuclear Research and consultancy Group, NL-1755ZG Petten (Netherlands); Deffrennes, Marc [European Commission, DG ENER, L-2530 Luxembourg (Luxembourg); Bredimas, Alexandre [LGI Consulting, 13 rue Marivaux, F-75002 Paris (France)

    2014-05-01

    This paper first provides an update on the current state of play and the potential future role of nuclear energy in Europe. It then describes the EU energy policy tools in the area of nuclear technology. It explains the three-tier strategy of the European nuclear technology platform and its demonstration initiatives, here specifically for nuclear cogeneration and HTR. The paper closes with an outlook on the boundary conditions at which HTR can become attractive for nuclear cogeneration, not only from an energy policy viewpoint but also economically.

  18. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  19. Chemisputtering of interstellar graphite grains

    International Nuclear Information System (INIS)

    Draine, B.T.

    1979-01-01

    The rate of erosion of interstellar graphite grains as a result of chemical reaction with H, N, and O is estimated using the available experiment evidence. It is argued that ''chemical sputtering'' yields for interstellar graphite grains will be much less than unity, contrary to earlier estimates by Barlow and Silk. Chemical sputtering of graphite grains in evolving H II regions is found to be unimportant, except in extremely compact (n/sub H/> or approx. =10 5 cm -3 ) H II regions. Alternative explanations are considered for the apparent weakness of the lambda=2175 A extinction ''bump'' in the direction of several early type stars

  20. Obtention of nuclear grade graphite

    International Nuclear Information System (INIS)

    Ferreira, M.L.

    1984-01-01

    The impurity level of natural graphite found in some of the most important mines of the State of Minas Gerais - Brasil is determined. It is also concerned with the development and use of natural graphite in nuclear reactors. Standard methods for chemical and instrumentsal analysis such as Spectrografic Determination by Emission, Spectrografic Determination by X-Rays, Spectrografic Determination by Atomic Asorption, Photometric Determination, and also chemical and physical methods for separation of impurities as well standard method for Estimating the Thermal Neutron Absorption Cross Section of graphite were employed. Some aditionals methods of purification to the ordinary treatment such as the use of metanol and halogens are also described. (Author) [pt