WorldWideScience

Sample records for fuel type guia

  1. Fuel loads and fuel type mapping

    Science.gov (United States)

    Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

    2003-01-01

    Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

  2. New type fuel exchange system

    International Nuclear Information System (INIS)

    Meshii, Toshio; Maita, Yasushi; Hirota, Koichi; Kamishima, Yoshio.

    1988-01-01

    When the reduction of the construction cost of FBRs is considered from the standpoint of the machinery and equipment, to make the size small and to heighten the efficiency are the assigned mission. In order to make a reactor vessel small, it is indispensable to decrease the size of the equipment for fuel exchange installed on the upper part of a core. Mitsubishi Heavy Industries Ltd. carried out the research on the development of a new type fuel exchange system. As for the fuel exchange system for FBRs, it is necessary to change the mode of fuel exchange from that of LWRs, such as handling in the presence of chemically active sodium and inert argon atmosphere covering it and handling under heavy shielding against high radiation. The fuel exchange system for FBRs is composed of a fuel exchanger which inserts, pulls out and transfers fuel and rotary plugs. The mechanism adopted for the new type fuel exchange system that Mitsubishi is developing is explained. The feasibility of the mechanism on the upper part of a core was investigated by water flow test, vibration test and buckling test. The design of the mechanism on the upper part of the core of a demonstration FBR was examined, and the new type fuel exchange system was sufficiently applicable. (Kako, I.)

  3. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  4. Inert matrix fuel in dispersion type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

    2006-06-30

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  5. Inert matrix fuel in dispersion type fuel elements

    Science.gov (United States)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  6. Guide for subdivision of spent fuel pool. Project UNESA MAAP5-SFP; Guia para subdivision de la piscina de combustible gastado. Proyecto UNESA MAAP5-SFP

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Barrios, M.; Garcia Gonzalez, M.; Perez Martin, F. J.

    2013-07-01

    The main goal of the UNESA MAAP5-SFP project is to analyze the capabilities of MAAP5 code and, particularly, the Spent Fuel Pool (SFP) module in order to tackle its modeling and facilitate the development of specific SFP models of Spanish NPPs. Within the project, Empresarios Agrupados (EEAA) is the responsible for the development of the Guide for the subdivision of the Spent Fuel Pool (SFP). This Guide includes a theoretical description of the model that is used by the code and a sequence of practical cases with the aim to evaluate the influence of specific parameters.

  7. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  8. A CAREM type fuel element dynamic analysis

    International Nuclear Information System (INIS)

    Magoia, J.E.

    1990-01-01

    A first analysis on the dynamic behaviour of a fuel element designed for the CAREM nuclear reactor (Central Argentina de Elementos Modulares) was performed. The model used to represent this dynamic behaviour was satisfactorily evaluated. Using primary estimations for some of its numerical parameters, a first approximation to its natural vibrational modes was obtained. Results obtained from fuel elements frequently used in nuclear power plants of the PWR (Pressurized Water Reactors) type, are compared with values resulting from similar analysis. (Author) [es

  9. Guias ópticos para sensores

    OpenAIRE

    Marcos de Castro Pacitti

    1993-01-01

    Neste trabalho foram estudados guias ópticos para atuarem como sensores ou em sistemas sensores. Ênfase especial foi dada em guias ópticos retangulares, para dispositivos ópticos integrados, e fibras ópticas elípticas. A analise de fibras ópticas elípticas resultou em um modelamento pratico para projeto das mesmas em aplicações onde se deseja operar com dois modos guiados, situação esta que se mostra muito atraente para implementação de diversos tipos de sensores a fibra óptica. A seguir es...

  10. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

    Aguilar H, F.; Mazon R, R.

    2001-05-01

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  11. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  12. Desempenho de modelos de calibração de guias de onda acopladas a TDR e a multiplexadores em três tipos de solos Performance of calibration models for TDR and multiplexer - connected waveguides in three soil types

    Directory of Open Access Journals (Sweden)

    Eugenio Ferreira Coelho

    2006-02-01

    Full Text Available O trabalho teve como objetivo avaliar modelos de calibração para dois tipos de guias de onda de TDR, referentes a dois equipamentos (Trase System e TDR 100, acopladas diretamente ao analisador de umidade ou a multiplexadores. Amostras de três tipos de solo foram acondicionadas em segmentos de tubos de PVC e saturadas. Dois tipos de guias de onda de três hastes, com capacitor e com resistor foram inseridas dentro de cada segmento de tubo com solo e conectadas a dois equipamentos de TDR, diretamente no testador de cabos ou via multiplexadores. Dados de umidade obtidos por gravimetria e da constante dielétrica foram tomados em cada coluna durante a secagem do solo da saturação até umidades próximas do limite inferior de disponibilidade de água por meio de leituras com as guias de onda conectadas ao testador de cabos e conectadas ao multiplexador. Um modelo polinomial cúbico foi ajustado aos dados da constante dielétrica do solo (épsilon e da correspondente umidade (teta e cinco modelos de determinação de q em função de e foram testados quanto ao desempenho. Os resultados mostraram que não houve diferença significativa na calibração das guias de onda com capacitor para uso com a TDR Trase System, considerando a conexão das guias ao analisador de umidade ou a multiplexadores. No caso da TDR 100, as guias de onda com resistor devem ser calibradas conforme o seu uso. O modelo cúbico foi o de melhor desempenho seguido pelo modelo de Roth que estimou, com boa exatidão, os valores da constante dielétrica e da umidade com a mais próximo de 0,5 para as guias de onda com capacitor que com as guias com resistor.The study aimed at evaluating of calibration models for two kinds of TDR waveguides used with Trase System and TDR 100 equipments linked to the cable tester or the multiplexer. Disturbed samples of three soils were packed in PVC columns and, after soil saturation, two TDR waveguides of three rods with capacitor and with

  13. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  14. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  15. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  16. The feasibility study on fuel types for the KALIMER

    International Nuclear Information System (INIS)

    Hwang, W.; Nam, C.; Yim, J. S.; Na, B. C.; Hahn, D. H.; Kim, Y. I.; Kim, Y. C.; Park, C. K.

    1997-08-01

    The economics of LMR is largely dependent on the construction cost of the power plant, and the fuel cycle options usually constitute 20 to 30 % of total electricity generation cost. The choice of fuel cycle technology and the fuel type is important in order to develop a LMR with better economics, performance and safety. The LMR fuel types, whose performances have been proven up to 15 at% burnup, are MOX and IFR metal fuel. The base alloy, binary (U-10% Zr) metal fuel with HT9 is used as structural materials of KALIMER. The design concept of KALIMER fuel has been established through the investigation of technical feasibilities on the fuel and recycle systems for MOX and IFR metal fuel. According to the results of comparative analysis for MOX and metal fuel, metal fuel is better than MOX in view of safety, in-reactor performance, nuclear characteristics, economics and non-proliferation, while MOX fuels have advantages in the developmental status and technical cooperation potential. The overall performance of binary (U-10% Zr) metal fuel with HT9 cladding, which is a potential start-up fuel for KALIMER, is not only superior to that of MOX fuel, but also has enough technical feasibility in its high-burnup performance, safety and economics. (author). 54 ref., 13 tabs., 20 figs

  17. Fuel exchange device for FBR type reactor

    International Nuclear Information System (INIS)

    Onuki, Koji.

    1993-01-01

    The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

  18. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  19. CANDU type fuel activities in Argentina

    International Nuclear Information System (INIS)

    Lavarez, L.; Casario, J.A.; Moreno, C.

    2003-01-01

    Domestic fuel performance in Embalse NPP during the last two years has been excellent without a significant occurrence of fuel failures. The defect rate level was reasonably low with a lowest value of 0.02 % in 2002. The implementation of fuel design optimizations to increase uranium content was fully completed by the end of year 2000. The in-reactor performance was not affected and shows the high degree of maturity reached for both the design and the manufacturing procedures and capabilities. A feasibility study for the utilization of SEU in Embalse NPP mainly conducted by NA-SA and AECL is almost completed. Some fuel related activities are still in progress. As part of them fuel behavior simulations using simplified power histories were performed to assess the influence of SEU fuel burnup extension. (author)

  20. Fuel cladding tube and fuel rod for BWR type reactor

    International Nuclear Information System (INIS)

    Urata, Megumu; Mitani, Shinji.

    1995-01-01

    A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

  1. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Kato, Shigeru.

    1993-01-01

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  2. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

    Moscalu, D.R.; Horhoianu, G.; Popescu, I.A.; Olteanu, G.

    1995-01-01

    In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

  3. Analyses for licensing of new fuel types at Paks NPP

    International Nuclear Information System (INIS)

    Kereszturi, A.; Bogatyr, S.; Miko, S.; Nemes, I.

    2003-01-01

    In the last years Paks NPP initiated several projects aiming at the introduction of new fuel types and resulting in more economic fuel cycles. The motivations, the reasons, and the economic consequences of the above modifications are detailed. The application of a new fuel type requires the renewal of the relevant chapters of the Safety Analysis Report. The fulfilment of fuel design basis requirements, to be summarised briefly also in the paper, must be investigated during normal and accidental conditions. The characteristics of the different codes, the data transfer between them are detailed. After, the cases of the Normal Operation, Anticipated Operation Occurrence, and the Postulated Accidents, judged as the most relevant ones in case of fuel modifications, are overviewed. In the last part, selected examples of the licensing calculations, performed by the above tools are presented. In conclusion, modifications of the WWER fuel, namely increased enrichment, application of burnable fuel pins, modified geometry make more economic fuel cycles (larger discharge burnup, power up-rate, reduced pressure vessel fluence) are possible. The further step (increased enrichment, burnable poison) of the fuel modernisation at NPP Paks is necessary for more economic fuel cycles and fuel consuming. A sound basis of licensing methodology, safety analysis, and necessary computer codes for the WWER fuel modernisation is available

  4. Fuel type characterization based on coarse resolution MODIS satellite data

    Directory of Open Access Journals (Sweden)

    Lanorte A

    2007-01-01

    Full Text Available Fuel types is one of the most important factors that should be taken into consideration for computing spatial fire hazard and risk and simulating fire growth and intensity across a landscape. In the present study, forest fuel mapping is considered from a remote sensing perspective. The purpose is to delineate forest types by exploring the use of coarse resolution satellite remote sensing MODIS imagery. In order to ascertain how well MODIS data can provide an exhaustive classification of fuel properties a sample area characterized by mixed vegetation covers and complex topography was analysed. The study area is located in the South of Italy. Fieldwork fuel type recognitions, performed before, after and during the acquisition of remote sensing MODIS data, were used as ground-truth dataset to assess the obtained results. The method comprised the following three steps: (I adaptation of Prometheus fuel types for obtaining a standardization system useful for remotely sensed classification of fuel types and properties in the considered Mediterranean ecosystems; (II model construction for the spectral characterization and mapping of fuel types based on two different approach, maximum likelihood (ML classification algorithm and spectral Mixture Analysis (MTMF; (III accuracy assessment for the performance evaluation based on the comparison of MODIS-based results with ground-truth. Results from our analyses showed that the use of remotely sensed MODIS data provided a valuable characterization and mapping of fuel types being that the achieved classification accuracy was higher than 73% for ML classifier and higher than 83% for MTMF.

  5. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto

    1990-01-01

    Various considerations are applied to fuel rods for improving the fuel burnup degree. If a gap between the fuel rods is changed, this varies the easiness for the flow of coolants depending on places, to reduce the thermal margin. Then, it is noted for the distribution of stresses generated due to the difference of water pressure caused by the difference of water streams between the inside and the outside of a channel box, and composite value, of stresses upon occurrence of earthquakes, neutron irradiation and a channel creep phenomenon caused by the stresses of due to the water pressure difference described above, the thickness of the channel box is increased in the upstream and decreased toward the downstream. Further, fuel spacers at the position where the thickness of the channel box is changed are spaced apart from the channel box so as not to brought into contact with the channel box. This can contribute to the reduction of coolants pressure loss, improvement of critical power and improvement of reactivity, as well as remarkably moderate local stresses applied from the fuel spacers to the channel box due to horizontal vibrations upon occurrence of earthquakes to improve the integrity of fuel assembly. (N.H.)

  6. Technical report: fabrication of PWR type rodlet fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Uno, Hisao; Sasajima, Hideo

    1990-06-01

    With respect to the simulated reactivity initiated accident (RIA) experiments with pre-irradiated LWR type fuel rods at nuclear safety research reactor (NSRR), there were principally three technical difficulties which should be overcome: (1) Fabrication of the rodlet fuel; Fuel rods from the commercial power reactors had an active column length by 3.6m. To utilize this for NSRR pulse experiment, rodlet fuel having an active column length by 0.12m (reduced to one thirtieth) is requested to fabricate without changing the inside fuel conditions. (2) Development of in-core instrumentations: During pre-irradiation stages, a long-sized fuel rod had dimensional changes by waterside corrosion, bowing, creep down and so on. The fuel also had greater amount of radioactive fission products. This condition is significant to in-core instrumentations to be attached to the fuel rods. Well characterized data to be obtained from these, however, are quite necessary and important from research point of view. Remote handling techniques to attach the rod pressure sensor, the cladding extensometer, the fuel extensometer, and the cladding surface thermocouple to pre-irradiated fuel rods are, therefore, requested to develop. (3) Installation of PIE equipments for pulsed rodlet fuels: PIE on the pulsed rodlet fuels are necessary to better understanding the fuel performance detaily. Equipments which can easily detect the data related to PCMI type fuel failure are matter of concern. Since 1986, the technical difficulties have been tried to overcome by all staffs belonging to Reactivity Accident Laboratory, NSRR Operation Division, Department of Reactor Fuel Examination and Hot Laboratory. This report describes the technical achievements obtained through four years work. (author)

  7. CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

  8. Fuel assemblies for FBR type reactor

    International Nuclear Information System (INIS)

    Ikeda, Kiyoshi.

    1981-01-01

    Purpose: To decrease errors in the flow rate distribution of coolants by resiliently inserting a flow regulation rod having a variable flow regulation element formed at the upper portion along the axial direction in the entrance nozzle of a fuel assembly. Constitution: A plurality of orifice aperture are formed to the entrance nozzle of a fuel assembly and an aperture for inserting a flow regulation rod is formed to the top end of the entrance nozzle. A fixed flow regulation element A and a variable flow regulation element B supported coaxially with the nozzle by a support ring are disposed to the inside of the nozzle. The element B is urged by the resilient urging spring to the element A and connected by way of support lever to the flow regulation rod. While on the other hand, the top end of the nozzle is inserted through the partition wall between a high pressure coolant chamber and a low pressure coolant chamber. An aperture for hydrodynamically supporting the fuel assembly is provided by way of a frame and a flow regulation rod that stands vertically from the low pressure coolant chamber is disposed to the center of the frame. In the fuel assembly, the flow regulation rod inserted from the aperture at the top end of the nozzle pushes the element B upwardly to thereby maintain a flow passage of the coolant between the elements A and B. (Seki, T.)

  9. Fuel saving type power plant for automobiles

    Energy Technology Data Exchange (ETDEWEB)

    Endo, N; Katsumoto, T; Shimizu, T; Hiramatsu, T; Fujita, Y

    1982-10-01

    Mitsubishi Motors Corporation has developed a modulated displacement engine named ''Orion MD'' and an electronically controlled damper clutch automatic transmission named ''ELC Automatic'' and has installed them on the new ''Mirage'' series and ''Cordia'' series, respectively, which were put on sale in February, 1982. They improve fuel economy to a great extent especially at low vehicle speed, and provide good driveability and high reliability. An outline of the ''Orion MD'' and ''ELC Automatic'' is presented.

  10. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Andrzejewski, Claudio de Sa

    2005-01-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  11. Effects of Fuel Type and Fuel Delivery System on Pollutant Emissions of Pride and Samand Vehicles

    Directory of Open Access Journals (Sweden)

    Akbar Sarhadi

    2017-04-01

    Full Text Available This research was aimed to study the effect of the type of fuel delivery system (petrol, dedicated or bifuel, the type of consumed fuel (petrol or gas, the portion of consumed fuel and also the duration of dual-fuelling in producing carbon monoxide, carbon dioxide and unburned hydrocarbons from Pride and Samand. According to research objectives, data gathering from 2000 vehicles has been done by visiting Hafiz Vehicle Inspection Center every day for 2 months. The results of this survey indicated that although there is no significant difference between various fuel delivery systems in terms of producing the carbon monoxide, carbon dioxide and unburned hydrocarbons by Samand, considering the emission amount of carbon dioxide, the engine performance of Pride in bifuel and dedicated state in GTXI and 132 types is more unsatisfactory than that of petrol state by 0.3 and 0.4%, respectively. On the other hand, consuming natural gas increases the amount of carbon monoxide emission in dual- fuel Pride by 0.18% and decreases that in dual-fuel Samand by 1.2%, which signifies the better design of Samand in terms of fuel pumps, used kit type and other engine parts to use this alternative fuel compared to Pride. Since the portion of consumed fuel and also duration of dual-fuelling does not have a significant effect on the amount of output pollutants from the studied vehicles, it can be claimed that the output substances from the vehicle exhaust are more related to the vehicle’s condition than the fuel type.

  12. Adesao ao guia alimentar para populacao brasileira

    Directory of Open Access Journals (Sweden)

    Eliseu Verly Junior

    2013-12-01

    Full Text Available OBJETIVO : Analisar a adesão ao Guia Alimentar para População Brasileira. MÉTODOS : Amostra composta por participantes do Inquérito de Saúde de São Paulo (n = 1.661 que preencheram dois recordatórios de 24 horas. Foi utilizado modelo bivariado de efeito misto para a razão entre o consumo de energia do grupo de alimentos e o consumo calórico total. A razão estimada foi utilizada para calcular o percentual de indivíduos com consumo abaixo ou acima da recomendação. RESULTADOS : Pelo menos 80,0% da população consome abaixo do recomendado para: leite e derivados; frutas e sucos de frutas; e cereais, tubérculos e raízes; aproximadamente 60,0% para legumes e verduras; 30,0% para feijões; e 8,0% para carnes e ovos. Adolescentes apresentaram a maior inadequação para legumes e verduras (90,0%, e o estrato de maior renda foi associado à menor inadequação para óleos, gorduras e sementes oleaginosas (57,0%. CONCLUSÕES : Foi observado consumo inadequado dos grupos de alimentos relacionados com aumento do risco de doenças crônicas.

  13. Econlit (EBSCOHost): guia d'ús. Gener 2011

    OpenAIRE

    Universitat de Barcelona. CRAI

    2011-01-01

    Guia de la base de dades Econlit sobre economia i finances produïda per l'American Economic Association (AEA) i distribuïda per EBSCOHost. És la font més important del món en referències a la literatura econòmica.

  14. Is fuel poverty in Ireland a distinct type of deprivation?

    OpenAIRE

    Watson, Dorothy; Maitre, Bertrand

    2014-01-01

    In this paper, we draw on the Central Statistics Office SILC data for Ireland to ask whether fuel poverty is a distinctive type of deprivation that warrants a fundamentally different policy response than poverty in general. We examine the overlap between fuel poverty (based on three self-report items) and poverty in general – with a particular emphasis on the national indicator of basic deprivation which is used in the measurement of poverty for policy purposes in Ireland. We examine changes ...

  15. Perfil vocal do guia de turismo Vocal profile of tourism guide

    Directory of Open Access Journals (Sweden)

    Elisângela Barros Soares

    2006-12-01

    Full Text Available OBJETIVO: caracterizar o perfil vocal dos guias de turismo, bem como gênero e idade. MÉTODOS: participaram desse estudo 23 guias de turismo, de ambos os gêneros, com idade entre 25 a 64 anos, participantes do Sindicato de Guias de Turismo do Estado de Pernambuco, que compareceram às reuniões trimestrais no período da coleta. Trata-se de um estudo de caráter descritivo, observacional e transversal. Para coleta foi realizada avaliação perceptivo-auditiva GRBAS. RESULTADOS: observou-se que a maioria dos guias apresentou loudness adequada, pitch normal e voz alterada. Além disso, as médias dos tempos máximos de fonação das vogais e das fricativas encontravam-se reduzidas e ataque vocal isocrônico. A ressonância, na maioria dos guias, estava equilibrada, mas houve uma incidência de ressonância laringo-faringea. A articulação foi precisa, com tipo e modo respiratório misto e nasal, respectivamente. Quanto à escala GRBAS as alterações apareceram de forma leve no G (grau de alteração vocal em 68%. CONCLUSÃO: na amostra estudada, a maioria era do gênero feminino com média de idade de 46 anos, e perfil vocal caracterizado por tempo máximo de fonação reduzidos, relação s/z adequado, ataque vocal isocrônico, pitch normal, loudness adequado, qualidade vocal alterada, com presença de rouquidão, soprosidade, tensão. A ressonância da maioria estava equilibrada e a articulação precisa, com tipo e modo respiratório misto e nasal, respectivamente. Quanto à escala GRBAS, as alterações apareceram de forma leve no grau de alteração vocal (G em 68% e tensão (S em 78% dos sujeitos.PURPOSE: to characterize the vocal profile of tourism guides, as well as gender and age. METHODS: 23 guides took part in this study, of both genders, with age between 25 to 64 years, partakers of the Union of Tourism Guides of the State of Pernambuco, who appeared to the quarterly meetings in the period of the collection. It is a descriptive

  16. Fuel assemblies for use in BWR type reactors

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1987-01-01

    Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

  17. Effect of engine parameters and gaseous fuel type on the cyclic variability of dual fuel engines

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed Y.E. Selim [United Arab Emirates University, Al-Ain (United Arab Emirates). Mechanical Engineering Department, Faculty of Engineering

    2005-05-01

    This paper presents an analysis of the cycle-to-cycle combustion variation as reflected in the combustion pressure data of a single cylinder, naturally aspirated, four stroke, Ricardo E6 engine converted to run as dual fuel engine on diesel and gaseous fuel of LPG or methane. A measuring set-up consisting of a piezo-electric pressure transducer with charge amplifier and fast data acquisition card installed on an IBM microcomputer was used to gather the data of up to 1200 consecutive combustion cycles of the cylinder under various combination of engine operating and design parameters. These parameters included type of gaseous fuel, engine load, compression ratio, pilot fuel injection timing, pilot fuel mass, and engine speed. The data for each operating conditions were analyzed for the maximum pressure, the maximum rate of pressure rise representing the combustion noise, and indicated mean effective pressure. The cycle-to-cycle variation is expressed as the mean value, standard deviation, and coefficient of variation of these three parameters. It was found that the type of gaseous fuel and engine operating and design parameters affected the combustion noise and its cyclic variation and these effects have been presented. 21 refs., 6 figs., 1 tab.

  18. Refining fuels of the heavy gas--oil type

    Energy Technology Data Exchange (ETDEWEB)

    Bruzac, J F.A.

    1930-01-28

    This invention has for its object the production of a new type of gas-oil fuel, obtained from crude petroleum, shale oil, and peat oil, according to the method of treatment mentioned, by means of which is obtained from gas oil, shale oil, lignite oil, and peat oil (deprived of asphaltic, and bituminous, resinous, and sulfur compounds), a fuel suitable for running Diesel, Junkers, and Clerget motors and all others of the same kind, by diminishing considerably the fouling and attack on the metal.

  19. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  20. Impact of Fuel Type on the Internal Combustion Engine Condition

    Directory of Open Access Journals (Sweden)

    Zdravko Schauperl

    2012-07-01

    Full Text Available The paper studies the influence of liquefied petroleum gas as alternative fuel on the condition of the internal combustion engine. The traffic, energy, economic and ecological influence as well as the types of fuel are studied and analyzed in an unbiased manner, objectively, and in detail, and the obtained results are compared with the condition of the engine of a vehicle powered by the stipulated fuel, petrol Eurosuper 95. The study was carried out on two identical passenger cars with one being fitted with gas installation. The obtained results show that properly installed gas installations in vehicles and the usage of LPG have no significant influence on the driving performances, but they affect significantly the ecological and economic parameters of using passenger cars.

  1. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  2. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  3. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

  4. Production of 15N for nitride type nuclear fuel

    International Nuclear Information System (INIS)

    Axente, Damian

    2005-01-01

    Full text: Nitride nuclear fuel is the choice for advanced nuclear reactors and ADS, considering its favorable properties as: melting point, excellent thermal conductivity, high fissile density, lower fission gas release and good radiation tolerance. The application of nitride fuels in different nuclear reactors requires use of 15 N enriched nitrogen to suppress 14 C production due to (n,p) reaction on 14 N. Nitride fuel is a promising candidate for transmutation in ADSs of radioactive minor actinides, which are converted into nitrides with 15 N for that purpose. Taking into account that at present the world wide 15 N market is about 20 - 40 Kg 15 N/y, the supply of that isotope for nitride type nuclear fuel, would demand an increase in production capacity by a factor of 1000. For an industrial plant producing 100 t/y 15 N at 99 at. % 15 N concentration, using present technology of 15 N/ 14 N isotopic exchange in Nitrox system, the first separation stage of the cascade would be fed with 10M HNO 3 solution at a 600 m 3 /h flow-rate. If conversion of HNO 3 into NO, NO 2 , at the enriching end of the columns, would be done with gaseous SO 2 , for an industrial plant of 100 t/y 15 N a consumption of 4 million t SO 2 /y and a production of 70 % H 2 SO 4 waste solution of 4.5 million m 3 /y are estimated. The reconversion of H 2 SO 4 into SO 2 in order to recycle SO 2 is a problem to be solved to compensate the cost of sulfur dioxide and to diminish the amount of sulfuric acid waste solution. It should be taken into consideration an important price reduction of 15 N in order to make possible its utilization for industrial production of nitride type nuclear fuel. (authors)

  5. Fabrication and characterization of MX-type fuels and fuel pins

    International Nuclear Information System (INIS)

    Richter, K.; Bartscher, W.; Benedict, U.; Gueugnon, J.F.; Kutter, H.; Sari, C.; Schmidt, H.E.

    1978-01-01

    This paper summarizes the most important fabrication parameters and characterization of fuel and fuel pins obtained during the investigation of uranium-plutonium carbides, oxicarbides, carbonitrides and nitrides in the past years at the European Institute for Transuranium Elements at Karlsruhe. All preparation methods discussed are based on carbothermic reduction of a mechanical blend of uranium-plutonium oxide and carbon powder. General data for carbothermic reduction processes are discussed (influence of starting material, homogeneity, control of degree of reaction, etc). A survey of different preparation methods investigated is given. Limitations with respect to temperature and atmosphere for both carbothermic reduction processes and sintering conditions for the different compounds are summarized. A special preparation process for mixed carbonitrides with low nitrogen content (U,Pu)sub(1-x)Nsub(x) in the range 0.1 0 C to 1400 0 C by means of a modulated electron beam technique. A scheme is proposed, which allows to predict the thermal properties of MX fuels on the basis of their chemical composition and porosity. Preparation, preirradiation characterization and final controls of fuel test pins for pellet and vibrocompacted type of pins are described and the most important data summarized for all advanced fuels irradiated at Dounreay (DN1) and Rapsodie Fast Reactor (DN2) within the TU irradiation programme

  6. Bi-fuel System - Gasoline/LPG in A Used 4-Stroke Motorcycle - Fuel Injection Type

    Science.gov (United States)

    Suthisripok, Tongchit; Phusakol, Nachaphat; Sawetkittirut, Nuttapol

    2017-10-01

    Bi-fuel-Gasoline/LPG system has been effectively and efficiently used in gasoline vehicles with less pollutants emission. The motorcycle tested was a used Honda AirBlade i110 - fuel injection type. A 3-litre LPG storage tank, an electronic fuel control unit, a 1-mm LPG injector and a regulator were securely installed. The converted motorcycle can be started with either gasoline or LPG. The safety relief valve was set below 48 kPa and over 110 kPa. The motorcycle was tuned at the relative rich air-fuel ratio (λ) of 0.85-0.90 to attain the best power output. From dynamometer tests over the speed range of 65-100 km/h, the average power output when fuelling LPG was 5.16 hp; dropped 3.9% from the use of gasoline91. The average LPG consumption rate from the city road test at the average speed of 60 km/h was 40.1 km/l, about 17.7% more. This corresponded to lower LPG’s energy density of about 16.2%. In emission, the CO and HC concentrations were 44.4% and 26.5% lower. Once a standard gas equipment set with ECU and LPG injector were securely installed and the engine was properly tuned up to suit LPG’s characteristics, the converted bi-fuel motorcycle offers efficiently, safely and economically performance with environmental friendly emission.

  7. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    Ranta-Puska, K.; Pihlatie, M.

    2001-01-01

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85 Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  8. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  9. Porosity in MX-type fuels and its stability

    International Nuclear Information System (INIS)

    Sari, C.

    1978-01-01

    Radial and axial temperature gradients were generated in MX-type fuels (U,Pu)C, (U,Pu)CN and (U,Pu)N in regions of temperature between 1000 and 2000 0 C. Typical temperature gradients were between 150 and 350 0 C/mm. Experiments show that under these conditions important restructuring of the fuel occurs after less than 40 hours. Densification in the thermal gradient was observed at temperature as low as 1100 0 C and the densification decreases with the increase of the nitrogen content. The grain growth rates decrease with the increase of the nitrogen content, thus paralleling the results of densification. Evidence of pore migration was found in the region with T approximately equal to 1500 0 C. Data of pore migration in MC and in carbon rich MCN plotted in an Arrhenius diagram gives a ΔH approximately equal to 95kcal/mole in approximate agreement with the values of evaporation enthalpy

  10. Status of fuel element technology for plate type dispersion fuels with high uranium density

    International Nuclear Information System (INIS)

    Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

    1983-01-01

    A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

  11. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  12. Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

    Science.gov (United States)

    Feddema, Rick

    Feddema, Rick T. M.S.M.E., Purdue University, December 2013. Effect of Aviation Fuel Type and Fuel Injection Conditions on the Spray Characteristics of Pressure Swirl and Hybrid Air Blast Fuel Injectors. Major Professor: Dr. Paul E. Sojka, School of Mechanical Engineering Spray performance of pressure swirl and hybrid air blast fuel injectors are central to combustion stability, combustor heat management, and pollutant formation in aviation gas turbine engines. Next generation aviation gas turbine engines will optimize spray atomization characteristics of the fuel injector in order to achieve engine efficiency and emissions requirements. Fuel injector spray atomization performance is affected by the type of fuel injector, fuel liquid properties, fuel injection pressure, fuel injection temperature, and ambient pressure. Performance of pressure swirl atomizer and hybrid air blast nozzle type fuel injectors are compared in this study. Aviation jet fuels, JP-8, Jet A, JP-5, and JP-10 and their effect on fuel injector performance is investigated. Fuel injector set conditions involving fuel injector pressure, fuel temperature and ambient pressure are varied in order to compare each fuel type. One objective of this thesis is to contribute spray patternation measurements to the body of existing drop size data in the literature. Fuel droplet size tends to increase with decreasing fuel injection pressure, decreasing fuel injection temperature and increasing ambient injection pressure. The differences between fuel types at particular set conditions occur due to differences in liquid properties between fuels. Liquid viscosity and surface tension are identified to be fuel-specific properties that affect the drop size of the fuel. An open aspect of current research that this paper addresses is how much the type of aviation jet fuel affects spray atomization characteristics. Conventional aviation fuel specifications are becoming more important with new interest in alternative

  13. Operating method of molten carbonate type fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, Tsuneo

    1988-12-06

    Molten carbonate type fuel cell involves a problem of oxidation of anode while the unit is stopped. Although there is a method proposed wherein an inactive gas is supplied to anode during the stoppage, the market-available inactive gas contains a slight amount of oxygen which makes it difficult to prevent the deterioration of the anode. In this invention, at the start and the stop other than the normal operation, a protective gas mixture of an inactive gas with a small amount of hydrogen is supplied to the anode. The inactive gas is a commercial type nitrogen, argon or helium; hydrogen is mixed in amount 0.5 - 2.0% of the inactive gas. By this method, oxygen in air which comes in from the gas-sealed portion of the cell is reduced by hydrogen in the protective gas and is discharged in the form of water. 2 figs.

  14. Thermal characteristics during hydrogen fueling process of type IV cylinder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Chan [Department of Fire and Disaster Prevention, Kyungil University, 33, Buhori, Hayang, Kyungsan 712-701 (Korea); Lee, Seung Hoon; Yoon, Kee Bong [Department of Mechanical Engineering, Chung Ang University, 221, Huksuk, Dongjak, Seoul 156-756 (Korea)

    2010-07-15

    Temperature increase during hydrogen fueling process is a significant safety concern of a high pressure hydrogen vessel. Hence, thermal characteristics of a Type IV cylinder during hydrogen filling process need to be understood. In this study, a series of experiments were conducted to quantify the temperature change of the cylinder during hydrogen filling to 35 MPa. Computational fluid dynamics (CFD) analysis was also conducted to simulate the conditions of the experiments. The results predicted by the CFD analysis show reasonable agreement with the experiments and the discrepancy between the CFD results and experimental results decrease with higher initial gas pressures. The upper and the lower parts of the vessel showed a temperature difference in the vertical direction. The upper gas temperature was higher than that of the lower part due to the buoyancy effect in the vessel. The maximum gas temperature was higher than the maximum temperature allowed in the ISO safety code (85 C) for the case in which the vessel was pressurized from 0 MPa to 35 MPa. This work contributes to the understanding of the thermal flow characteristics of the hydrogen filling process and notes that additional efforts should be made to guarantee the safety of a type IV cylinder during the hydrogen fueling process. (author)

  15. Guia de receitas brasileiras: uma saborosa viagem pela literatura

    OpenAIRE

    Santos, Denísia Moraes dos

    2011-01-01

    Esta dissertação tem por objetivo compreender, sob o ponto de vista da Análise Dialógica do Discurso, o processo de construção dos sentidos em Guia de receitas brasileiras ─ texto premiado na primeira edição do Concurso Cultural Viagem Nestlé pela Literatura (1999). Com o tema A literatura e a vida nos 500 anos de Brasil , essa edição desafiou alunos e professores a produzir um intertexto literário-cultural, utilizando cinco obras da literatura brasileira ─ Cronistas do Descobrim...

  16. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

    1987-01-01

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  17. 14 CFR 26.33 - Holders of type certificates: Fuel tank flammability.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Holders of type certificates: Fuel tank... Tank Flammability § 26.33 Holders of type certificates: Fuel tank flammability. (a) Applicability. This... part 25 of this chapter. (2) Exception. This paragraph (b) does not apply to— (i) Fuel tanks for which...

  18. The prospects of use of alternative types of fuel in road transport ...

    African Journals Online (AJOL)

    The article is devoted to the analysis of possibilities of using alternative types of fuel in transport. Gas engine fuels are considered as potential energy carriers for diesel engines. Since the constructions of vehicles, using gas and traditional types of fuel, have some differences, the most important are the issues of ensuring ...

  19. 14 CFR 26.37 - Pending type certification projects: Fuel tank flammability.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Pending type certification projects: Fuel tank flammability. 26.37 Section 26.37 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... AIRPLANES Fuel Tank Flammability § 26.37 Pending type certification projects: Fuel tank flammability. (a...

  20. Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

    2009-01-01

    A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

  1. Study of homogeneous fuel cells type 10 x 10

    International Nuclear Information System (INIS)

    Montes, J.L.; Perusquia, R.; Ortiz, J.J.; Francois, J.L.; Marquez, C.M.

    2005-01-01

    At the moment in the National Institute of Nuclear Research (ININ) are carried out studies with the purpose of to establish a methodology that allows to carry out the neutron design of fuel cells of type 10 x 10. During the initial stage of the process of cells design, starting from the data that have to do with the planned energy demand it requires to be estimated the average value of the enrichment in U 235 w/o of the one assemble. The experience has shown that the accuracy that is achieved in this estimate it depends, among other factors, of the information (e.g. concentrations of U 235 and Gd 2 O 3 ) of the cells that its are disposed in that moment. For what we consider convenient to enlarge the available information by means of a series of calculations of cell physics; and to the one same time some aspects can be studied on the parameters that define the characteristics of a fuel cell. In this work the effect of the presence of different distributions of the concentrations of the fissile material is analyzed and of burnup poisons on the reactivity parameters of the cell as well as in the peak factor of local power (LPPF-Local Power Peaking Factor). (Author)

  2. Design and analytic evaluation of a rim effect reduction type LWR fuel for extending burnup

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo; Kameyama, Takanori; Kinoshita, Motoyasu

    1991-01-01

    We have designed a new concept fuel design 'Rim effect reduction type fuel' which has thin natural UO 2 layer on surface of a UO2 pellet. Our neutronic analyses with ANRB code show this fuel design can reduce rim effect (burnup at plelet rim) by about 30 GWd/t comparing a normal fuel. It is known that a high burnup fuel has different microstructure from as-fabricated one at fuel rim (which is called as rim region) due to rim effect. Therefore this fuel design can expect smaller rim region than a normal fuel. Our fuel performance analyses with EIMUS code show this fuel design can reduce fuel center temperature at high burnup if thermal conductivity of fuel pellet decreases with burnup in inverse proportion. However, this fuel design increases fuel center temperature at low and middle burnup than a normal fuel due to increase of thermal power density at pellet center. Additionally Irradiation experiment of this fuel design can be considered to offer important data which make clear the relation between rim effect and fuel performance. (author)

  3. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

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

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  4. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  5. Dissolution process for advanced-PWR-type fuels

    International Nuclear Information System (INIS)

    Black, D.E.; Decker, L.A.; Pearson, L.G.

    1979-01-01

    The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

  6. Fuel rod for use in BWR type reactor

    International Nuclear Information System (INIS)

    Takeuchi, Kiyoshi.

    1989-01-01

    A hollow intermediate end plug is disposed to a plenum portion of a fuel rod and a plenum spring is disposed between the end plug and the upper end of a fuel pellet. Then, a hollow portion is disposed between the intermediate end plug and an upper end plug. Thus, since a only a non exothermic portion is present from the intermediate end plug to the upper end plug, oxidation, corrosion, etc. to the fuel can are not caused so much as in the exothermic portion. Accordingly, the wall thickness of the fuel may be reduced to such a extent as only capable of withstanding the external pressure by coolants and the increasing inner pressure due to the release of FP gases and, accordingly, the wall thickness can be reduced as compared with that of the fuel portion in the fuel can. Further, since the power density per unit length of the fuel rod is reduced for fuels with increased number of fuel rods, it is possible to design so as to reduce the release amount of FP gases thereby decreasing the plenum volume. Further, since the surface area in the coolant phase stream portion is reduced, it can be expected for decreasing the pressure loss of fuels and accompanying effect for improving the channel stability. (T.M.)

  7. Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

    1996-01-01

    In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

  8. Fuel consumption models for pine flatwoods fuel types in the southeastern United States

    Science.gov (United States)

    Clinton S. Wright

    2013-01-01

    Modeling fire effects, including terrestrial and atmospheric carbon fluxes and pollutant emissions during wildland fires, requires accurate predictions of fuel consumption. Empirical models were developed for predicting fuel consumption from fuel and environmental measurements on a series of operational prescribed fires in pine flatwoods ecosystems in the southeastern...

  9. Fuel assembly for use in BWR type reactor

    International Nuclear Information System (INIS)

    Inaba, Yuzo.

    1988-01-01

    Purpose: To attain the reduction of neutron irradiation amount to control rods by the improvement in the reactor shutdown margin and the improvement of the control rod worth, by enhancing the arrangement of burnable poisons. Constitution: The number of burnable poison-incorporated fuel rods present in the outer two rows along the sides in adjacent with a control rod among the square lattice arrangement in a fuel assembly is decreased to less than 1/4 for that of total burnable poison-incorporated fuel rods, while the remaining burnable posion-incorporated fuel rods are arranged in the region other than above (that is, those regions not nearer to the control rod). Thus, even if a sufficient number of burnable poison to prolong the controlling effect for the reactivity with the burnable contents as the fuel assembly are disposed, only the burnable poison -incorporated fuel rods by the number less than 1/4 for that of the total burnable poison-incorporated fuel rods are present near the control rod of the fuel assembly. Accordingly, the control rod worth at the initial stage of the burning is increased at both high and normal temperatures. (Kawakami, Y.)

  10. Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

    2009-06-15

    Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

  11. Assessment of fuel damage of pool type research reactor in the case of fuel plates blockage

    Energy Technology Data Exchange (ETDEWEB)

    Jalil, Jafari; Samad, Khakshournia [AEOI, Karegar Ave. School of R and D of Nuclear Reactors and Accelerators, Teheran (Iran, Islamic Republic of); D' Auria, F. [Pisa Univ., DIMNP (Italy)

    2007-07-01

    Tehran Research Reactor (TRR) is a pool type 5 MW research reactor. It is assumed that external objects or debris that may fall down to reactor core cause obstruction of coolant flow through one of the fuel assemblies. Thermal hydraulic analysis of this event, using the RELAP5 system code has been studied. The reported transient is related to the partial and total obstruction of a single Fuel Element (FE) cooling channel of 27 FE equilibrium core of TRR. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FE integrity. Two scenarios are analysed to emphasize the severity of the accident. The first one is a partial blockage of an average FE considering four different obstruction levels: 25%, 50%, 75% and 97% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FE. This study constitutes the first step of a larger work which consists of performing a 3-dimensional simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation included in the RELAP5 code. Main results obtained from the RELAP5 calculations are as following. First, in the case of flow blockage under 97% of the nominal flow area of an average FE, only an increase of the coolant and clad temperatures is observed without any consequences for the integrity of the FE. The mass flow rate remains sufficient to cool the clad safely. Secondly, in the case of total obstruction of the nominal flow area, it is seen that transient turns out to be a severe accident due to the dryout conditions are reached shortly and melting of the cladding occurs. Thirdly, the use of the point kinetic approach leads to conservative results. A best estimate simulation of such kind of transients requires the use of 3-dimensional kinetic calculations, which could be done using the current Coupled Codes

  12. Field experience of new nuclear fuel types on the Kola NPP

    International Nuclear Information System (INIS)

    Adeev, V.; Burlov, S.; Panov, A.; Saprykin, V.

    2008-01-01

    Specificity of the Kola nuclear power plant geographical position, conditions of region economics determine fuel management strategy. Isolation of Kola power supply system and, as a consequence, generating capacities redundancy cause operation of the nuclear power plant on reduced power level. At the same time there is a need to operate the power unit on the maximum power level in the case of not planned conditions. The basis of in-core fuel management is an achievement of the maximal burnup under providing of high installed capacity. At present there are not abilities to improve the fuel cycle based on traditional implementation fuel assemblies. Burnup maximum in these fuel cycles is achieved. At the core periphery installed highest possible quantity of the burned-up assemblies in the view of safety operation margins satisfaction. Works on application of the second generation fuel have been carried out on the Kola NPP since 2002. Fuel assemblies of this type are profiled. Burnable absorber, changed lattice spacing in relation to standard fuel, changed height of a fuel column, thickness of fuel pin clad are applied. In CR fuel followers modernized docking unit (with hafnium plates are intended for energy-release splash suppression) is used. At present 2-nd generation fuel is in experimental operation on unit 3 (18-21 fuel cycles, 2002-2007 years) and unit 4 (18-19 fuel cycles, 2005-2007 years). Safety margins did not exceeded. Coolant activity did not exceed the limiting value. There were not damaged fuel assemblies of second generation. Originally in the project of applications of new fuel it was supposed to refuel annually 78 fresh assemblies. At the moment annual refueling consists of 66 assemblies with effective enrichment 3.82 %. Cycle duration does not exceed 250-260 effective days. The part of assemblies is left on 5-th cycle of operation. In a similar fuel cycle in 2007 on the unit 1 operation with profiled fuel (enrichment of 3.82 %) of shakeproof type

  13. Fast reactor fuel pin behavior analyses in a LOF type transient event

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

    2013-06-01

    In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

  14. Measurement of the Velocity and Pressure Drop in a Tubular Type Fuel

    International Nuclear Information System (INIS)

    Jonghark Park; Heetaek Chae; Cheol Park; Heonil Kim

    2006-01-01

    We have developed a tubular type fuel assembly design as one of candidates for fuel to be used in the Advanced HANARO Reactor (AHR). The tubular type fuel has several merits over a rod type fuel with respect to the thermal-hydraulic and structural safety; the larger ratio of surface area to volume makes the surface temperature of a fuel element become lower, and curved plate is stronger against longitudinal bending and vibration. In the other side, a disadvantage is expected such that the flow velocity can be distributed unevenly channel by channel because the flow channels are isolated from each other in a tubular type fuel assembly. In addition to the design development, we also investigated the flow characteristics of the tubular fuel experimentally. To examine the flow velocity distribution and pressure drop, we made an experiment facility and a mockup of the tubular fuel assembly. The fuel assembly consists of 6 concentric fuel tubes so that 7 layers are made between fuel tubes. Since each layer is divided into three sections by stiffeners, 21 isolated flow channels are made in total. We employed pitot-tubes to measure the coolant velocity in each channel. The maximum velocity was measured as large as about 28% of the average velocity. It was observed in the innermost channel contrarily to the expectation from the hydraulic diameter. A change in the total flow rate did not affect the flow distribution. Meanwhile, the pressure drop was measured as about 70% of the drop in the rod type fuel assembly in use in HANARO. (authors)

  15. Rearrangement of fuel assemblies in the RBMK type reactors to flatten power distribution and improve the fuel cycle

    International Nuclear Information System (INIS)

    Mityaev, Yu.I.; Vikulov, V.K.

    1982-01-01

    A possibility of increasing the burnup of uranium fuel unloaded from the RBMK type reactors is investigated. Three variants of a two-zone reactor-refueling are considered: 1. the simplest variant of continuous refueling used at present, when the central and peripherical reactor zones are additionally fueled independently by similar fuel assemblies (FA); 2. the variant under which new FA are loaded to the peripherical zone and are used there up to the same burnup as in the first case, then all the peripherical FA (PFA) are rearranged to the centre and they are used there up to maximum burnup; 3. the same as in the second variant, but not all the PFA are rearranged to the centre but only FA with small fuel burnup. It is shown by calculation that average fuel burnup for the third refueling variant is several per cent higher at the optimal burnup of rearranged FA. Besides, flattening of fuel channel power is improved in this case, that permits to increase uranium enrichment and burnup at the same maximum power. It essentially improves economic parameters of the reactor. It is concluded that realization of the considered variant of fuel refueling will produce the most essential effect for reactors refueled without shutdown

  16. Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Kim, Dong Min; Lee, Yoon Sang; Cheong, Yong Moo

    2011-01-01

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

  17. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    International Nuclear Information System (INIS)

    Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

    2004-01-01

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

  18. A review of microstructural analysis on U3Si2-Al plate-type fuel

    International Nuclear Information System (INIS)

    Ti Zhongxin; Guo Yibai

    1995-12-01

    The microstructure of U 3 Si 2 -Al plate-type fuel, that is the microstructure of fuel particles, compatibility of the fuel particles and Al matrix, fuel particles distribution, dogbone area morphology, clad and meat thickness, bone quality of clad/frame and clad/fuel core, and the effect of these factors on products quality were comprehensively investigated and analyzed by means of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD), energy dispersive X-ray spectrometry (EDX), image processing technique, etc.. The main results are as following: U-7.7%Si alloy contains two phases: primary U 3 Si 2 and small amount of USi (about 12%), free-uranium was not detected in fuel particles; the dogbone area is the key factor affecting fuel plate quality (1 ref., 16 figs., 4 tabs.)

  19. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  20. Fuel assemblies for use in FBR type reactors

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo; Terasawa, Michitaka.

    1984-01-01

    Purpose: To prevent slackings in lapping wires and thereby enabling to maintain the distance between fuel pins always constant during use. Constitution: Lapping wires are wound helically around the outer circumference of each fuel pin in order to maintain the distance between fuel pins constant and unify the flow of coolants. The material of the lapping wire is defined as below. Specifically, austenite stainless steels incorporated with 0.045% titanium are used in the state of molten procession material as they are without no further cold working. Lapping wires having anti-swelling property can be obtained with this material and the slackings in the lapping wires during use can be prevented. (Ikeda, J.)

  1. In-use vs. type-approval fuel consumption of current passenger cars in Europe

    International Nuclear Information System (INIS)

    Ntziachristos, L.; Mellios, G.; Tsokolis, D.; Keller, M.; Hausberger, S.; Ligterink, N.E.; Dilara, P.

    2014-01-01

    In-use fuel consumption data of 924 passenger cars (611 petrol, 313 diesel) were collected from various European sources and were evaluated in comparison to their corresponding type-approval values. The analysis indicated that the average in-use fuel consumption was higher than the type-approval one by 11% for petrol cars and 16% for diesel cars. Comparison of this dataset with the Travelcard database in the Netherlands showed that the deviation increased for late model years and in particular for cars with low type-approval values. The deviation was higher than 60% for vehicles registered in 2012 within the 90–100 gCO 2 /km bin. Unrealistic vehicle resistances used in type-approval were identified as one of the prime reasons of the difference. A simplified linear model developed in the study may be used to predict in-use fuel consumption based on data publicly available. The model utilizes the fuel consumption measured in type-approval, the mass, and the engine capacity to provide in-use fuel consumption. This may be either used to correct fuel consumption factors currently utilized by emission models (e.g. COPERT, HBEFA, VERSIT+, and others) or could be used independently to make projections on how fuel consumption may develop on the basis of changing future passenger cars characteristics. - Highlights: • In-use fuel consumption of petrol and diesel passenger cars is 11% and 16% higher than type-approval, respectively. • The relative difference between in-use and type-approval increases for late model and vehicles with low consumption. • Unrealistically low vehicle resistances are identified as a prime reason of low type-approval fuel consumption. • A model developed predicts in-use consumption on the basis of type-approval consumption, vehicle mass, and engine capacity

  2. Modeling solid-fuel dispersal during slow loss-of-flow-type transients

    International Nuclear Information System (INIS)

    DiMelfi, R.J.; Fenske, G.R.

    1981-01-01

    The dispersal, under certain accident conditions, of solid particles of fast-reactor fuel is examined in this paper. In particular, we explore the possibility that solid-fuel fragmentation and dispersal can be driven by expanding fission gas, during a slow LOF-type accident. The consequences of fragmentation are studied in terms of the size and speed of dispersed particles, and the overall quantity of fuel moved. (orig.)

  3. The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

    International Nuclear Information System (INIS)

    Caligara, F.; Campana, M.; Mandler, R.; Blank, H.

    1979-01-01

    Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well

  4. Feasibility of Electromagnetic Acoustic Evaluation for Quality Test of a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Lee, Yoon Sang; Cheong, Yong Moo

    2010-01-01

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel core in aluminum alloy. Recently KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done with water, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined within this paper is a non-ferromagnetic material such as aluminum which has a good acousto-elastic property, for an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an EMAT technology for an automated inspection of a nuclear fuel without water

  5. Crossflow characteristics of flange type fuel element for very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Kaburaki, Hideo; Suzuki, Kunihiko; Nakamura, Masahide.

    1987-01-01

    Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element. (author)

  6. Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

    International Nuclear Information System (INIS)

    Zhang Zhuohua; Yu Junchong; Peng Shinian

    2014-01-01

    According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

  7. Development of a 200kW multi-fuel type PAFC power plant

    Energy Technology Data Exchange (ETDEWEB)

    Take, Tetsuo; Kuwata, Yutaka; Adachi, Masahito; Ogata, Tsutomu [NTT Integrated Information & Energy System Labs., Tokyo (Japan)

    1996-12-31

    Nippon Telegraph and Telephone Corporation (NFT) has been developing a 200 kW multi-fuel type PAFC power plant which can generate AC 200 kW of constant power by switching fuel from pipeline town gas to liquefied propane gas (LPG) and vice versa. This paper describes the outline of the demonstration test plant and test results of its fundamental characteristics.

  8. Influence of the silicon content on the core corrosion properties of dispersion type fuel plates

    International Nuclear Information System (INIS)

    Calvo, C.; Saenz de Tejada, L. M.; Diaz Diaz, J.

    1969-01-01

    A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI 3 and AI 2 O 3 according to the reaction. (Author)

  9. Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

    International Nuclear Information System (INIS)

    Sonoda, Yukio; Tamaoki, Tetsuo

    1989-01-01

    Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

  10. Progress of the Russian RERTR program: Development of new-type fuel elements for Russian-built research reactors

    International Nuclear Information System (INIS)

    Vatulin, A. V.; Stetskiy, Y.A.; Mishunin, V.A.; Suprun, V.B.; Dobrikova, I.V.

    2002-01-01

    The new design of pin-type fuel elements and fuel assembly on their basis for Russian research reactors has been developed. The number of following activities has been performed: computational and experimental substantiation of fuel element design; development of fabrication process of fuel elements; manufacturing of experimental assembly for lifetime in-pile tests. The relevant fuel assemblies are considered to be perspective for usage as low-enriched fuel for Russian research reactors. (author)

  11. Cladding tube of fuel rod for a BWR type reactor

    International Nuclear Information System (INIS)

    Nakayama, Hitoshi; Fujie, Kunio; Kuwahara, Heikichi; Hirai, Tadamasa; Kakizaki, Kimio.

    1976-01-01

    Object: To form a cladding tube wall with tunnels in communication with the exterior through a number of small-diameter openings to rapidly disperse a large quantity of heat thereby providing high density of the fuel rod. Structure: Tunnels adjacent to each other are provided under the skin in contact with cooling liquid of a cladding tube, and a number of openings through which said tunnels and the periphery of the cladding tube are placed in communication are formed, said openings each having its section smaller than that of said tunnel. With this arrangement, the cooling water entered the tunnel through some of small diameter openings absorbs heat of the fuel rod to be vaporized, which is flown out into the cooling water through the other small diameter openings and formed into vapor bubbles which move up for release of heat. (Taniai, N.)

  12. Pyrochlore-type catalysts for the reforming of hydrocarbon fuels

    Science.gov (United States)

    Berry, David A [Morgantown, WV; Shekhawat, Dushyant [Morgantown, WV; Haynes, Daniel [Morgantown, WV; Smith, Mark [Morgantown, WV; Spivey, James J [Baton Rouge, LA

    2012-03-13

    A method of catalytically reforming a reactant gas mixture using a pyrochlore catalyst material comprised of one or more pyrochlores having the composition A.sub.2-w-xA'.sub.wA''.sub.xB.sub.2-y-zB'.sub.yB''.sub.zO.sub.7-.DELTA.. Distribution of catalytically active metals throughout the structure at the B site creates an active and well dispersed metal locked into place in the crystal structure. This greatly reduces the metal sintering that typically occurs on supported catalysts used in reforming reactions, and reduces deactivation by sulfur and carbon. Further, oxygen mobility may also be enhanced by elemental exchange of promoters at sites in the pyrochlore. The pyrochlore catalyst material may be utilized in catalytic reforming reactions for the conversion of hydrocarbon fuels into synthesis gas (H.sub.2+CO) for fuel cells, among other uses.

  13. Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

    International Nuclear Information System (INIS)

    Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

    2011-01-01

    A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

  14. Design verification testing for fuel element type CAREM

    International Nuclear Information System (INIS)

    Martin Ghiselli, A.; Bonifacio Pulido, K.; Villabrille, G.; Rozembaum, I.

    2013-01-01

    The hydraulic and hydrodynamic characterization tests are part of the design verification process of a nuclear fuel element prototype and its components. These tests are performed in a low pressure and temperature facility. The tests requires the definition of the simulation parameters for setting the test conditions, the results evaluation to feedback mathematical models, extrapolated the results to reactor conditions and finally to decide the acceptability of the tested prototype. (author)

  15. Vehicle type choice under the influence of a tax reform and rising fuel prices

    DEFF Research Database (Denmark)

    Mabit, Stefan Lindhard

    2014-01-01

    change in new vehicle purchases toward more diesel vehicles and more fuel-efficient vehicles. The paper analyses to what extent a vehicle tax reform similar to the Danish 2007 reform may explain changes in purchasing behaviour. The paper investigates the effects of a tax reform, fuel price changes......, and technological development on vehicle type choice using a mixed logit model. The model allows a simulation of the effect of car price changes that resemble those induced by the tax reform. This effect is compared to the effects of fuel price changes and technology improvements. The simulations show...... that the effect of the tax reform on fuel efficiency is similar to the effect of rising fuel prices while the effect of technological development is much larger. The conclusion is that while the tax reform appeared in the same year as a large increase in fuel efficiency, it seems likely that it only explains...

  16. R and D activities on CANDU-type fuel in Indonesia

    International Nuclear Information System (INIS)

    Suripto, A.; Badruzzaman, M.; Latief, A.

    1997-01-01

    The status of R and D activities in Indonesia with respect of CANDU-type fuel development is presented. The activities have been started since the first feasibility study to introduce nuclear power plants was carried out in 1970s. The early research comprised the in-situ pilot production of yellow-cake in Kalimantan (Borneo) experimental mining site, uranium purification and pellet preparation. This program continued to gain a full support from the Government which culminated in the realisation of the construction by BATAN of a large fuel development laboratory in Serpong, starting from 1984 in co-operation with NIRA Ansaldo of Italy. The laboratory, which is called the Power Reactor Experimental Fuel Element Installation (EFEI) was originally designed as an experimental facility to integrate the acquired domestic R and D results gained so far on the CANDU-type fuel technology and the additional know-how received from NIRA Ansaldo which at that time was engaged, in developing a CANDU-type fuel, called the CIRENE fuel design. In the present days the facility houses the power reactor fuel development activities carried out to build up the national capability on power reactor fuel fabrication technology in anticipation to embark upon the nuclear energy era in the near future. (author)

  17. Stressed and strained state for cermetic-rod-type fuel element

    International Nuclear Information System (INIS)

    Kulikov, I.S.

    1987-01-01

    Calculation technique for designing the stress-strained state of a cermetic rod-type fuel element has been proposed. The technique is based on the time-dependent step-by-step method and the solution of the deformation equilibrium equation for continuous and thick-wall long cylinders at every temporal step by the finite difference method. Additional strains, caused by thermal expansion and radiation swelling, have been taken into account. The transion from the non-contact model to the stiff-contact model has been provided in the case of cladding-fuel gap dissappearing in one or a number of cross-sections along the fuel element height. The method is supplemented by the formula for fuel cans stability estimation in the case of high coolant external pressure. The example of estimation of the cermetic-rod-type fuel elements are considered as an example

  18. Investigation and proposal of the system to affect nuclear fuel type authorization and analysis code certification

    International Nuclear Information System (INIS)

    2006-03-01

    In order to develop the system to affect more advanced and rational regulations of nuclear fuels and earlier introduction of new technologies in nuclear power plants, domestic and overseas safety regulation systems and state of their implementation for water cooled reactor fuel and safety analysis code had been investigated and new regulation system to affect nuclear fuel type authorization and analysis code certification was proposed. Topical report system for common parts related with nuclear fuel type authorization and analysis code certification was firstly proposed for knowledge base. Maintaining consistent safety examination supported by experts, introduction of domestic efficient system for lead irradiation test fuel, and analysis code certification and quality assurance were also proposed. (T. Tanaka)

  19. Tema 8. Principis físics dels semiconductors (Guia del tema)

    OpenAIRE

    Beléndez Vázquez, Augusto

    2011-01-01

    Guia del "Tema 8. Principis físics dels semiconductors" de l'assignatura "Fonaments Físics de l'Enginyeria I" de "Grau en Enginyeria en So i Imatge" impartit a l'Escola Politècnica Superior de la Universitat d'Alacant.

  20. Study on new-type fuel-related assembly handling tools for PWR NPP

    International Nuclear Information System (INIS)

    Fan Xiumei

    2013-01-01

    This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

  1. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    International Nuclear Information System (INIS)

    Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

    1977-01-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  2. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

    1977-04-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  3. Fuel type characterization and potential fire behavior estimation in Sardinia and Corsica islands

    Science.gov (United States)

    Bacciu, V.; Pellizzaro, G.; Santoni, P.; Arca, B.; Ventura, A.; Salis, M.; Barboni, T.; Leroy, V.; Cancellieri, D.; Leoni, E.; Ferrat, L.; Perez, Y.; Duce, P.; Spano, D.

    2012-04-01

    Wildland fires represent a serious threat to forests and wooded areas of the Mediterranean Basin. As recorded by the European Commission (2009), during the last decade Southern Countries have experienced an annual average of about 50,000 forest fires and about 470,000 burned hectares. The factor that can be directly manipulated in order to minimize fire intensity and reduce other fire impacts, such as three mortality, smoke emission, and soil erosion, is wildland fuel. Fuel characteristics, such as vegetation cover, type, humidity status, and biomass and necromass loading are critical variables in affecting wildland fire occurrence, contributing to the spread, intensity, and severity of fires. Therefore, the availability of accurate fuel data at different spatial and temporal scales is needed for fire management applications, including fire behavior and danger prediction, fire fighting, fire effects simulation, and ecosystem simulation modeling. In this context, the main aims of our work are to describe the vegetation parameters involved in combustion processes and develop fire behavior fuel maps. The overall work plan is based firstly on the identification and description of the different fuel types mainly affected by fire occurrence in Sardinia (Italy) and Corsica (France) Islands, and secondly on the clusterization of the selected fuel types in relation to their potential fire behavior. In the first part of the work, the available time series of fire event perimeters and the land use map data were analyzed with the purpose of identifying the main land use types affected by fires. Thus, field sampling sites were randomly identified on the selected vegetation types and several fuel variables were collected (live and dead fuel load partitioned following Deeming et al., (1977), depth of fuel layer, plant cover, surface area-to-volume ratio, heat content). In the second part of the work, the potential fire behavior for every experimental site was simulated using

  4. Fuels planning: science synthesis and integration; social issues fact sheet 02: Developing personal responsibility for fuels reduction: Types of information to encourage proactive behavior

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    Fuels management responsibilities may include providing local property owners with the information for taking responsibility for reducing fuels on their land. This fact sheet discusses three different types of information that may be useful in programs to engage property owners in fuel reduction activities.

  5. Development of core technology for research reactors using plate type fuels

    International Nuclear Information System (INIS)

    Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

    2009-12-01

    Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

  6. Performance evaluation of CPF shredder type mechanical crusher with simulated core fuel pin

    International Nuclear Information System (INIS)

    Nakahara, Masaumi; Sano, Yuichi; Aose, Shin-ichi

    2006-12-01

    In the advanced aqueous reprocessing system, powder fuel dissolution has been investigated, which is quite effective on the dissolution for highly concentrated solution. As one of the effective means that powder the irradiated MOX fuel, we have been developing shredder type mechanical crusher. This apparatus can automatically crush the sheared fuel pieces by twin-shaft disk blades, powder the crushed fragments by disk blades and screen blade, and recover the powdered fuel. The shredder type mechanical crusher was developed for using in a hot cell in Chemical Processing Facility, and the first crush experiment with this crusher was carried out at July 2004 using the simulated core fuel pin. This experiment showed that the crushed fragments could not be grinded efficiency because screen blade vibrated up and down during the operation. Additionally, the strength of screen blade block was insufficient to crush the sheared fuel pieces stably. Therefore, about 70% of fuel was recovered in maximum. Based on the results of the first experiment, screen blade was fixed up mainly and the second experiment was carried out with improved apparatus at September 2005. In this experiment, about 96% of fuel could be recovered in maximum because screen blade was stable during the operation. (J.P.N.)

  7. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  8. Perspectives for practical application of the combined fuel kernels in VVER-type reactors

    International Nuclear Information System (INIS)

    Baranov, V.; Ternovykh, M.; Tikhomirov, G.; Khlunov, A.; Tenishev, A.; Kurina, I.

    2011-01-01

    The paper considers the main physical processes that take place in fuel kernels under real operation conditions of VVER-type reactors. Main attention is given to the effects induced by combinations of layers with different physical properties inside of fuel kernels on these physical processes. Basic neutron-physical characteristics were calculated for some combined fuel kernels in fuel rods of VVER-type reactors. There are many goals in development of the combined fuel kernels, and these goals define selecting the combinations and compositions of radial layers inside of the kernels. For example, the slower formation of the rim-layer on outer surface of the kernels made of enriched uranium dioxide can be achieved by introduction of inner layer made of natural or depleted uranium dioxide. Other potential goals (lower temperature in the kernel center, better conditions for burn-up of neutron poisons, better retention of toxic materials) could be reached by other combinations of fuel compositions in central and peripheral zones of the fuel kernels. Also, the paper presents the results obtained in experimental manufacturing of the combined fuel pellets. (authors)

  9. Finite element modelling of different CANDU fuel bundle types in various refuelling conditions

    International Nuclear Information System (INIS)

    Roman, M. R.; Ionescu, D. V.; Olteanu, G.; Florea, S.; Radut, A. C.

    2016-01-01

    The objective of this paper is to develop a finite element model for static strength analysis of the CANDU standard with 37 elements fuel bundle and the SEU43 with 43 elements fuel bundle design for various refuelling conditions. The computer code, ANSYS7.1, is used to simulate the axial compression in CANDU type fuel bundles subject to hydraulic drag loads, deflection of fuel elements, stresses and displacements in the end plates. Two possible situations for the fuelling machine side stops are considered in our analyses, as follows: the last fuel bundle is supported by the two side stops and a side stop can be blocked therefore, the last fuel bundle is supported by only one side stop. The results of the analyses performed are briefly presented and also illustrated in a graphical form. The finite element model developed in present study is verified against test results for endplate displacement and element bowing obtained from strength tests with fuel bundle string and fuelling machine side-stop simulators. Comparison of ANSYS model predictions with these experimental results led to a very good agreement. Despite the difference in hydraulic load between SEU43 and CANDU standard fuel bundles strings, the maximum stress in the SEU43 endplate is about the same with the maximum stress in the CANDU standard endplate. The comparative assessment reveals that SEU43 fuel bundle is able to withstand high flow rate without showing a significant geometric instability. (authors)

  10. BH2201 type leakage monitoring equipment of reactor fuel elements

    International Nuclear Information System (INIS)

    Ji Changsong; Dai Zhude; Xie Liangnian; Zhang Shulan; Zhang Shuheng

    1999-01-01

    A high-sensitive equipment monitoring leakage of the reactor fuel elements has been developed. The delayed neutrons emitted from fission product-pioneer nucleus are monitored in the 1st circle water. An array of 3 He proportional counter tubes is designed as a neutron detector for delayed neutrons, the detection geometry of which is near to 4π. In order to reduce the influence of interference factors the monitoring of fission product is carried out with 75s delay. The 87 Br delayed neutron pioneer nucleus is chosen as a monitoring object. The neutron detection efficiency of the developed equipment is 6.1%, which is 3 times higher than one of all available advanced equipment of the same function both at home and abroad

  11. Measurements and observations on microscopic swelling in MX-type fuels

    International Nuclear Information System (INIS)

    Ronchi, C.; Ray, I.L.F.; Thiele, H.; Laar, J. van de.

    1978-01-01

    Microscopic swelling has been investigated by electron microscopy in several MX-type fuels, irradiated in fast and thermal neutron flux. The results show that fission gas bubbles in these compounds grow to large sizes if the in-pile fuel temperature rises above a critical value (swelling critical temperature Tsub(C)). A comparison has been made of the swelling rates in fuels of different composition, showing that Tsub(C) increases from carbides to nitrides. In fuels subjected to in-pile restructuring (highly rated) He-bonded pins microscopic swelling is affected by pore and grain boundary migration. The influence of these phenomena on the fuel swelling performance has been discussed

  12. Evaluation of plate type fuel elements by eddy current test method

    International Nuclear Information System (INIS)

    Frade, Rangel Teixeira

    2015-01-01

    Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

  13. Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel

    International Nuclear Information System (INIS)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

    2009-01-01

    An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

  14. Power distribution gradients in WWER type cores and fuel failure root causes

    Energy Technology Data Exchange (ETDEWEB)

    Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

    2014-02-15

    Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

  15. Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

    International Nuclear Information System (INIS)

    Ravnik, M.

    1988-11-01

    The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

  16. Power release estimation inside of fuel pins neighbouring fuel pin with gadolinium in a WWER-1000 type core

    International Nuclear Information System (INIS)

    Mikus, J.

    2006-01-01

    The purpose of this work consists in investigation of the gadolinium fuel pin (fps) influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of neighbouring FPs that could result in static loads with some consequences, e.g., FP bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, relevant information about power release inside of needed (investigated) FP, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized in a WWER-1000 type core with 7 fuel assemblies at zero boron concentration and containing gadolinium FPs. Application of the above evaluation method is demonstrated on investigated FP neighbouring a FP with gadolinium by means of the 1) Azimuthal power distribution inside of investigated FP on their fuel pellet surface in horizontal plane and 2) Gradient of the power distribution inside of investigated FP in two opposite positions on pellets surface that are situated to- and outwards a FP with gadolinium. Similar information can be relevant from the viewpoint of the FP failures occurrence investigation (Authors)

  17. Studying some regimes of the WWER-440 type reactor failed fuel element operation

    International Nuclear Information System (INIS)

    Aksenov, N.A.; Samsonov, B.V.; Sulaberidze, V.Sh.; Frej, A.K.

    1981-01-01

    The results of investigating the serviceability of experimental fuel elements close by type to that of the WWER-440 type reactor in the cans of which untightness in the form of small opening are made. The tests are carried out in the SM-2 reactor high temperature water loop at the temperature of 473 K, pressure of (1-2)x10 4 kPa, coolant flow rate of 3.7-5.5 m 3 /h. The analysis of the obtained results shows that the character of changes in the fission product (FP) activity in the circuit in a considerable extent is determined bt the thermal-optical conditions of the fuel element operation. If water in the gap between fuel and can does not boil, activity changes smoothly and bursts caused by increased FP release are observed only under transient conditions of reactor operation. In the presence of water boiling in the gap the FP release has of impulse character with the frequency determined besides the untightness dimension by free volume inside the fuel element can (with its increase the pulsation frequency increases). FP release from fuel is connected with their direct escape from an open surface. When water in the gap the FP release from the fuel element occurs practically immediately. Without boiling the FP delay in the gap is determined by their diffusion in a layer of water. The conclusion is drawn that the FP release from failed fuel elements may be reduced by eliminating the water boiling in the gap between the fuel and the can by means of the fuel element power or coolant temperature decrease

  18. Enrichment measurement in TRIGA type fuels; Medicion de enriquecimiento en combustibles tipo Triga

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-05-15

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  19. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  20. Core conversion from rod to plate type fuel elements in research reactors

    International Nuclear Information System (INIS)

    Khattab, M.S.; Mina, A.R.

    1997-01-01

    Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

  1. Loss-of-flow test L5 on FFTF-type irradiated fuel

    International Nuclear Information System (INIS)

    Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

    1978-03-01

    Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O 2 fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times

  2. Development of simulation code for FBR spent fuel dissolution with rotary drum type continuous dissolver

    International Nuclear Information System (INIS)

    Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi

    2011-01-01

    Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO 2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (∼500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified. (author)

  3. Development of new type concrete for spent fuel storage cask

    International Nuclear Information System (INIS)

    Shimojo, J.; Mantani, K.; Owaki, E.; Sugihara, Y.; Hata, A.; Shimono, M.; Taniuchi, H.

    2004-01-01

    Heat resistant concrete has been developed to make it possible to design a new type cask that has been designed on the same concept of metal cask technologies for use in high temperature conditions. The allowable temperature of conventional concrete is limited to less than 100 degrees Celsius because most of its moisture is free water and therefore hydrogen, which is effective for neutron shielding, can be easily lost. Our newly developed concrete uses chemically bonded water and as a result can be used under high temperatures

  4. Development of new type concrete for spent fuel storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Shimojo, J.; Mantani, K. [Kobe Steel, Ltd., Hyogo (Japan); Owaki, E.; Sugihara, Y.; Hata, A.; Shimono, M. [Taisei Corp., Tokyo (Japan); Taniuchi, H. [Transnuclear, Ltd., Tokyo (Japan)

    2004-07-01

    Heat resistant concrete has been developed to make it possible to design a new type cask that has been designed on the same concept of metal cask technologies for use in high temperature conditions. The allowable temperature of conventional concrete is limited to less than 100 degrees Celsius because most of its moisture is free water and therefore hydrogen, which is effective for neutron shielding, can be easily lost. Our newly developed concrete uses chemically bonded water and as a result can be used under high temperatures.

  5. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

  6. Evidence of asymmetric behavioral responses to changes in gasoline prices and taxes for different fuel types

    International Nuclear Information System (INIS)

    Bajo-Buenestado, Raúl

    2016-01-01

    Using monthly data from the Spanish gasoline retail market we explore asymmetries in consumers’ behavioral responses to changes in gasoline prices and taxes. In particular, we are interested in investigating whether an increase in gasoline taxes has a more negative impact on the demand than a –similar in magnitude– increase in the “pre-tax” price of gasoline for different fuel types. We estimate fuel consumers’ responses using a rich set of robust panel data models considering potential dynamic effects and endogeneity problems. We find evidence to confirm the existence of asymmetric responses for the demand of unleaded fuels and agricultural diesel fuel. However we cannot support this statement for the regular diesel case: for this fuel both the tax-exclusive price and the tax elasticities are roughly the same. This result agrees with the fact that “diesel drivers” tend to be better informed about changes in both fuel prices and taxes. Some implications in terms of fiscal policy and pollution and climate change policy are also discussed. - Highlights: •Provide evidence of asymmetric responses of gasoline demand due to changes in prices and taxes. •Identify differences in the elasticity of the demand of diesel fuel and unleaded gasoline. •Perform robustness checks considering dynamic effects and IV regression. •Provide some policy recommendations for future gasoline tax changes.

  7. Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

    International Nuclear Information System (INIS)

    Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

    2008-01-01

    Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

  8. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  9. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  10. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  11. A dynamic, dependent type system for nuclear fuel cycle code generation

    Energy Technology Data Exchange (ETDEWEB)

    Scopatz, A. [The University of Chicago 5754 S. Ellis Ave, Chicago, IL 60637 (United States)

    2013-07-01

    The nuclear fuel cycle may be interpreted as a network or graph, thus allowing methods from formal graph theory to be used. Nodes are often idealized as nuclear fuel cycle facilities (reactors, enrichment cascades, deep geologic repositories). With the advent of modern object-oriented programming languages - and fuel cycle simulators implemented in these languages - it is natural to define a class hierarchy of facility types. Bright is a quasi-static simulator, meaning that the number of material passes through a facility is tracked rather than natural time. Bright is implemented as a C++ library that models many canonical components such as reactors, storage facilities, and more. Cyclus is a discrete time simulator, meaning that natural time is tracked through out the simulation. Therefore a robust, dependent type system was developed to enable inter-operability between Bright and Cyclus. This system is capable of representing any fuel cycle facility. Types declared in this system can then be used to automatically generate code which binds a facility implementation to a simulator front end. Facility model wrappers may be used either internally to a fuel cycle simulator or as a mechanism for inter-operating multiple simulators. While such a tool has many potential use cases it has two main purposes: enabling easy performance of code-to-code comparisons and the verification and the validation of user input.

  12. Food Science and Technology Abstracts (FSTA): guia ràpida. Gener 2011

    OpenAIRE

    Universitat de Barcelona. CRAI

    2011-01-01

    Guia ràpida de la base de dades bibliogràfica (FSTA) d'àmbit mundial sobre ciència, tecnologia i química alimentària, nutrició i salut humana, biotecnologia i toxicologia. Buida prop de 1800 revistes especialitzades, monografies, conferències, tesis, patents, legislació, etc. publicats en unes 40 llengües.

  13. 14 CFR 291.44 - BTS Schedule P-12(a), Fuel Consumption by Type of Service and Entity.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false BTS Schedule P-12(a), Fuel Consumption by... TRANSPORTATION Reporting Rules § 291.44 BTS Schedule P-12(a), Fuel Consumption by Type of Service and Entity. (a.... (e)(1) The cost of fuel shall include shrinkage, but excludes: (i) “Throughput” and “in to plane...

  14. 40 CFR 600.209-08 - Calculation of vehicle-specific 5-cycle fuel economy values for a model type.

    Science.gov (United States)

    2010-07-01

    ...-cycle fuel economy values for a model type. 600.209-08 Section 600.209-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year Automobiles-Procedures for...

  15. Influence of the voids fraction in the power distribution for two different types of fuel assemblies

    International Nuclear Information System (INIS)

    Jacinto C, S.; Del Valle G, E.; Alonso V, G.; Martinez C, E.

    2017-09-01

    In this work an analysis of the influence of the voids fraction in the power distribution was carried out, in order to understand more about the fission process and the energy produced by the fuel assembly type BWR. The fast neutron flux was analyzed considering neutrons with energies between 0.625 eV and 10 MeV. Subsequently, the thermal neutron flux analysis was carried out in a range between 0.005 eV and 0.625 eV. Likewise, its possible implications in the power distribution of the fuel cell were also analyzed. These analyzes were carried out for different void fraction values: 0.2, 0.4 and 0.8. The variations in different burn steps were also studied: 20, 40 and 60 Mwd / kg. These values were studied in two different types of fuel cells: Ge-12 and SVEA-96, with an average initial enrichment of 4.11%. (Author)

  16. A summary report on recruitment type researches on nuclear fuel cycle in fiscal year of 2001

    International Nuclear Information System (INIS)

    2002-07-01

    The promotion system on recruitment type researches on nuclear fuel cycle begun on fiscal year of 1999, aims to intend to activate researching environment of the Japan Nuclear Cycle Development Institute (JNC) through intercourses, information exchanges, publication of research results, and so on between researchers in other organizations and JNC, as a result, to effectively promote fundamental and basic R and Ds. This report contains summaries of 28 items of research results on the recruitment type researches on nuclear fuel cycle as 9 items relating to fast breeder reactors, 8 items relating to nuclear fuel cycle, 1 item relating to radiation safety, and 10 items relating to geological disposal and science, carried out on fiscal year of 2001. (G.K.)

  17. Effect of engine parameters and type of gaseous fuel on the performance of dual-fuel gas diesel engines. A critical review

    Energy Technology Data Exchange (ETDEWEB)

    Sahoo, B.B. [Centre for Energy, Indian Institute of Technology, Guwahati 781039 (India); Sahoo, N.; Saha, U.K. [Department of Mechanical Engineering, Indian Institute of Technology, Guwahati 781039 (India)

    2009-08-15

    Petroleum resources are finite and, therefore, search for their alternative non-petroleum fuels for internal combustion engines is continuing all over the world. Moreover gases emitted by petroleum fuel driven vehicles have an adverse effect on the environment and human health. There is universal acceptance of the need to reduce such emissions. Towards this, scientists have proposed various solutions for diesel engines, one of which is the use of gaseous fuels as a supplement for liquid diesel fuel. These engines, which use conventional diesel fuel and gaseous fuel, are referred to as 'dual-fuel engines'. Natural gas and bio-derived gas appear more attractive alternative fuels for dual-fuel engines in view of their friendly environmental nature. In the gas-fumigated dual-fuel engine, the primary fuel is mixed outside the cylinder before it is inducted into the cylinder. A pilot quantity of liquid fuel is injected towards the end of the compression stroke to initiate combustion. When considering a gaseous fuel for use in existing diesel engines, a number of issues which include, the effects of engine operating and design parameters, and type of gaseous fuel, on the performance of the dual-fuel engines, are important. This paper reviews the research on above issues carried out by various scientists in different diesel engines. This paper touches upon performance, combustion and emission characteristics of dual-fuel engines which use natural gas, biogas, producer gas, methane, liquefied petroleum gas, propane, etc. as gaseous fuel. It reveals that 'dual-fuel concept' is a promising technique for controlling both NO{sub x} and soot emissions even on existing diesel engine. But, HC, CO emissions and 'bsfc' are higher for part load gas diesel engine operations. Thermal efficiency of dual-fuel engines improve either with increased engine speed, or with advanced injection timings, or with increased amount of pilot fuel. The ignition

  18. Radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: professional exposure during mormal operation

    International Nuclear Information System (INIS)

    White, I.F.; Kelly, G.N.

    1983-01-01

    The radiological impact of the fuel cycle of light water type reactors using enriched uranium may be changed by plutonium recycle. The impact on human population and on the persons professionally exposed may be different according to the different steps of the fuel cycle. This report analyses the differential radiological impact on the different types of personnel involed in the fuel cycle. Each step of the fuel cycle is separately studied (fuel fabrication, reactor operation, fuel reprocessing), as also the transport of the radioactive materials between the different steps. For the whole fuel cycle, one estimates that, with regard to the fuel cycle using enriched uranium, the plutonium recycle involves a small increase of the professional exposure

  19. The Tourist Attractions in Curitiba - PR: An overview by Guia Brazil Quatro Rodas

    Directory of Open Access Journals (Sweden)

    Franciele Cristina Manosso

    2015-04-01

    Full Text Available The tour guides books can be worked as relevant instruments for the dissemination of destinations, because encompass information about the destinations' attractions, equipment, infrastructure and tourism services. Then, the tourist can through this, conceiving a destination’s image that will visit, because with the elements worked by the guides, they can get a comprehensive view of what is offered by the destination. Therefore, this article has as objective the analysis of the tourist attractions in Curitiba’s city, according to Guia Brazil Quatro Rodas, presented in the years 1989, 2001, 2013. Thus, the paper aims to also highlight how the document presents the city to its visitors, analyzing the description and the photographs that are showing Curitiba to the tourist. For this, was used as methodology the literature's and documents' research, the latter being applied to observe which attractions, description, and photos were used in the dissemination of Curitiba in the Guia Brazil. As the study results, it can be noted that there was an increase in the number of tourist attractions in Curitiba presented by Guia Brazil Quatro Rodas, and some of these have progressed in the respect of their classification, in the other words, they have improved in quality. Another issue is the publication’s evolution in question the destinations’ presentation to the reader, because more information came to be published, and also began to be used photos to show to the tourist the Curitiba destination.

  20. The effects of alternative fuel types on the organoleptic qualities of ...

    African Journals Online (AJOL)

    This study was carried out to assess the effects of alternative smoking fuel types on the organoleptic qualities of coarse pork sausages. The sausages were produced with lean pork (2.5 kg) and pork fat (0.5 kg), minced, mixed with spices and stuffed into natural casings. They were grouped into four and each group was ...

  1. Heat conduction in a plate-type fuel element with time-dependent boundary conditions

    International Nuclear Information System (INIS)

    Faya, A.J.G.; Maiorino, J.R.

    1981-01-01

    A method for the solution of boundary-value problems with variable boundary conditions is applied to solve a heat conduction problem in a plate-type fuel element with time dependent film coefficient. The numerical results show the feasibility of the method in the solution of this class of problems. (Author) [pt

  2. Contact-type displacement measuring mechanism for fuel assembly in reactor

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Ko, Kuniaki.

    1995-01-01

    The measuring mechanism of the present invention, which is used in a lmfbr type reactor, is suspended by a gripper of a fuel handing machine, and it comprises a combination of a displacement amount measuring jig allowed to be inserted into a handling head of a fuel assembly and a displacement amount measuring ring disposed at the lower portion in the handling head. The displacement amount measuring jig has a structure comprising a releasable handle and a columnar or cylindrical measuring portion allowable to be inserted into the handling head formed at the lower portion of the handle, which are connected with each other. When an interference (contact) occurred between the displacement amount measuring jig and the stepwise displacement amount measuring ring during the measurement, change of load and a phenomenon that the fuel handing machine can not be lowered are recognized, so that core displacement amount can be recognized based on the stroke of the gripper portion. Then, remote measurement is possible for displacement and deformation of the fuel assembly in the reactor container, and the measurement can be conducted by the same procedures and in the same period of time as in a case of ordinary fuel exchange operation. A flow channel for coolants passing through the fuel assembly can be ensured, thereby enabling to measure the amount of core displacement which is closer to an actual value in the reactor. (N.H.)

  3. Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

    International Nuclear Information System (INIS)

    Adlys, G.; Adliene, D.

    2002-01-01

    Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

  4. Fuel loading method to exchangeable reactor core of BWR type reactor and its core

    International Nuclear Information System (INIS)

    Koguchi, Kazushige.

    1995-01-01

    In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

  5. Investigations on burning efficiency and exhaust emission of in-line type emulsified fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Tseng, Y.K. [National Chinyi University of Technology (Taiwan). Dept. of Mechanical Engineering; Cheng, H.C. [Point Environmental Protection Technology Company Limited (Taiwan)

    2011-07-28

    In this research, the burning efficiency as well as exhaust emission of a new water-in-oil emulsified fuel system was studied. This emulsified system contains two core processes, the first one is to mix 97% water with 3% emulsifier by volume, and get the milk-like emulsified liquid, while the second one is to compound the milk-like emulsified liquid with heavy oil then obtain the emulsified fuel. In order to overcome the used demulsification problem during in reserve or in transport, this system was designed as a made and use in-line type. From the results of a series of burning tests, the fuel saving can be 8--15%. Also, from the comparison of decline for the heat value and total energy output of emulsified fuel, one can find that the water as the dispersed phase in the combustion process will lead to a micro-explosion as well as the water gas effect, both can raise the combustion temperature and burning efficiency. By comparing the waste gas emission of different types of emulsified fuel, one can know that, the CO2 emission reduces approximately 14%, and NOx emission reduces above 46%, meaning the reduction of the exhaust gas is truly effective. From the exhaust temperature of tail pipe, the waste heat discharge also may reduce 27%, it is quite advantageous to the global warming as well as earth environmental protection.

  6. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

    2016-10-21

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

  7. The influence of the types of marine fuel over the Energy Efficiency Operational Index

    Science.gov (United States)

    Acomi, Nicoleta; Acomi, Ovidiu

    2014-05-01

    One of the main concerns of our society is certainly the environment protection. The international efforts for maintaining the environment clean are various and this paper refers to the efforts in the maritime transport field. Marine pollution consists of the water pollution and also the air pollution. Regardless of the delay in recognizing the later type of pollution, it rapidly gains many organizations to argue on it. The first step was including a dedicated annex (Annex VI) in the International Convention for the Prevention of Pollution from Ships, in 1997, which seeks to minimize the airborne emissions from ships. In order to control and minimize the air pollution, the International Maritime Organization has also developed a series of measures for monitoring the emissions. These measures are grouped in three main directions: technical, operational and management related. The subject of our study is the concept of Energy Efficiency Operational Index (EEOI), developed to provide ship-owners with assistance in the process of establishing the emissions from ships in operation, and to suggest the methods for achieving their reduction. As a monitoring tool, EEOI represents the mass of CO2 emitted per unit of transport work. The actual CO2 emission from combustion of fuel on board a ship during each voyage is calculated by multiplying total fuel consumption for each type of fuel (e.g. diesel oil, gas oil, light fuel oil, heavy fuel oil, liquefied petroleum gas, liquefied natural gas) with the carbon to CO2 conversion factor for the fuel in question. The performed transport work is calculated by multiplying mass of cargo (tonnes, number of TEU/cars, or number of passengers) with the distance in nautical miles corresponding to the transport work done. Using the software developed by the author it will be emphasized the variation of the EEOI value for one vessel using different types of fuel for the voyage's legs (distance to discharge port, distance to loading port, the

  8. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  9. Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

    International Nuclear Information System (INIS)

    Fleming Rubio, Peter Alex

    2010-01-01

    The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

  10. Method of fabricating zirconium metal for use in composite type fuel cans

    International Nuclear Information System (INIS)

    Imahashi, Hiromichi; Inagaki, Masatoshi; Akabori, Kimihiko; Tada, Naofumi; Yasuda, Tetsuro.

    1985-01-01

    Purpose: To mass produce zirconium metal for fuel cans with less radiation hardening. Method: Zirconium sponges as raw material are inserted in a hearth mold and a procedure of melting the zirconium sponges portionwise by using a melting furnace having electron beams as a heat source while moving the hearth is repeated at least for once. Then, the rod-like ingot after melting is melted again in a vacuum or inert gas atmosphere into an ingot of a low oxygen density capable of fabrication. A composite fuel can billet is formed by using the thus obtained zirconium ingot and a zircalloy, and a predetermined composite type fuel can is manufactured by way of hot extrusion and pipe drawing fabrication. The raw material usable herein is zirconium sponge with an oxygen density of 400 ppm or higher and the content of impurity other than oxygen is between 1000 - 5000 ppm in total, or the molten material thereof. (Kamimura, M.)

  11. Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel

    International Nuclear Information System (INIS)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

    2010-01-01

    A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

  12. Experimental study on the 300W class planar type solid oxide fuel cell stack: Investigation for appropriate fuel provision control and the transient capability of the cell performance

    International Nuclear Information System (INIS)

    Komatsu, Y; Brus, G; Szmyd, J S; Kimijima, S

    2012-01-01

    The present paper reports the experimental study on the dynamic behavior of a solid oxide fuel cell (SOFC). The cell stack consists of planar type cells with standard power output 300W. A Major subject of the present study is characterization of the transient response to the electric current change, assuming load-following operation. The present studies particularly focus on fuel provision control to the load change. Optimized fuel provision improves power generation efficiency. However, the capability of SOFC must be restricted by a few operative parameters. Fuel utilization factor, which is defined as the ratio of the consumed fuel to the supplied fuel is adopted for a reference in the control scheme. The fuel flow rate was regulated to keep the fuel utilization at 50%, 60% and 70% during the current ramping. Lower voltage was observed with the higher fuel utilization, but achieved efficiency was higher. The appropriate mass flow control is required not to violate the voltage transient behavior. Appropriate fuel flow manipulation can contribute to moderate the overshoot on the voltage that may appear to the current change. The overshoot on the voltage response resulted from the gradual temperature behavior in the SOFC stack module.

  13. Experimental study on the 300W class planar type solid oxide fuel cell stack: Investigation for appropriate fuel provision control and the transient capability of the cell performance

    Science.gov (United States)

    Komatsu, Y.; Brus, G.; Kimijima, S.; Szmyd, J. S.

    2012-11-01

    The present paper reports the experimental study on the dynamic behavior of a solid oxide fuel cell (SOFC). The cell stack consists of planar type cells with standard power output 300W. A Major subject of the present study is characterization of the transient response to the electric current change, assuming load-following operation. The present studies particularly focus on fuel provision control to the load change. Optimized fuel provision improves power generation efficiency. However, the capability of SOFC must be restricted by a few operative parameters. Fuel utilization factor, which is defined as the ratio of the consumed fuel to the supplied fuel is adopted for a reference in the control scheme. The fuel flow rate was regulated to keep the fuel utilization at 50%, 60% and 70% during the current ramping. Lower voltage was observed with the higher fuel utilization, but achieved efficiency was higher. The appropriate mass flow control is required not to violate the voltage transient behavior. Appropriate fuel flow manipulation can contribute to moderate the overshoot on the voltage that may appear to the current change. The overshoot on the voltage response resulted from the gradual temperature behavior in the SOFC stack module.

  14. Dietary recommendations: comparing dietary guidelines from Brazil and the United States Recomendações dietéticas: comparação entre os guias alimentares brasileiro e americano

    Directory of Open Access Journals (Sweden)

    Rosely Sichieri

    2010-11-01

    Full Text Available The Brazilian dietary guidelines are based in part on mainstream United States' recommendations, in spite of the criticisms and shortcomings of the American guidelines. In this paper, Brazilian food guidelines are summarized and discussed in comparison with the USA recommendations. American and Brazilian dietary recommendations are quite similar in many aspects, particularly those related to variety in the diet, the importance of physical activity and weight management. Different to American guidelines, those from Brazil advise people to choose fresh foods, to prefer healthier types of fat, to limit trans fat intake and to eat good sources of protein, but does not recommend the consumption of whole grains. Besides the challenges related to their implementation, indicators for the evaluation of the effectiveness of these guidelines should be established from the beginning, particularly those related to changes in dietary habits and the prevalence of obesity.O guia alimentar brasileiro é baseado parcialmente nas recomendações americanas a despeito das críticas e problemas identificados no documento dos Estados Unidos. Neste artigo, as recomendações alimentares para o Brasil são resumidas e discutidas em comparação com as recomendações estadunidenses. Os guias alimentares brasileiro e americano são bastante similares em diversos aspectos, particularmente aqueles relacionados com a variação da dieta, a importância da atividade física e o gerenciamento do peso. Diferentemente dos Estados Unidos, o guia brasileiro estimula o consumo de alimentos frescos, aconselha o uso de fontes saudáveis de gorduras, a limitação do consumo de gordura trans, o consumo de boas fontes de proteínas, mas não indica o consumo de grãos integrais. Além dos desafios relacionados com a sua implantação, os indicadores para a avaliação da eficácia dos guias alimentares devem ser estabelecidos desde sua implantação, particularmente, os relacionados

  15. Preliminary study of cost benefits associated with duplex fuel pellets of the LOWI type

    International Nuclear Information System (INIS)

    Ainscough, J.B.; Coucill, D.N.; Howl, D.A.; Jensen, A.; Misfeldt, I.

    1983-01-01

    Duplex UO 2 pellets, which consist of an outer enriched annulus and a depleted or natural core, can provide a solution to the problem of stress corrosion cracking failures, which have led to constraints being placed on ramp rates in power reactors. An analysis of the reactor physics and the performance of duplex pellets is presented in the context of a 17 X 17 pressurized water reactor fuel rod design. The study has been based on the particular type of duplex pellet in which the core and the annulus are physically separate; this is called ''LOWI'' after the Danish design. At low burnup, this fuel shows a significant improvement in power ramp performance compared with standard fuel. At higher burnup, the benefits are less certain but as the severity of the ramp will usually be less in high burnup fuel simply because of the reduced rating, the reduction in benefit may not be significant. If the gap between the core and annulus persists to high burnup, there will be no loss of benefit. Economic calculations and a cost-benefit analysis are presented to show the number of extra full-power hours of reactor operation that must be obtained in order to outweigh the additional fabrication costs associated with this fuel

  16. Calculational modeling of fuel assemblies of WWER-1000 type with the use of burnable absorber Gadolinum; comparative analysis

    International Nuclear Information System (INIS)

    Yeremenko, M.L.; Kovbasenko, Yu.P.; Loetsch, T.

    2001-01-01

    In connection with the beginning of the use of fuel assemblies with burnable absorbers by integration of Gadolinum into the nuclear fuel at Ukrainian NPP the task of testing the code systems and the pertinent neutron cross section libraries for the new fuel arose. Taking into account the long term experience of German experts with calculations and evaluation of nuclear fuel containing Gadolinum it was decided to carry out a series of test calculations for fuel assembly lattices of PWR, WWER-440 and WWER-1000 types using the NESSEL/PYTHIA and CASMO/SIMULATE code systems (Authors)

  17. Re-qualification of MTR-type fuel plates fabrication process

    International Nuclear Information System (INIS)

    Elseaidy, I.M.; Ghoneim, M.M.

    2010-01-01

    The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)

  18. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  19. Fuel pin bowing and related investigation of the gadolinium fuel pin influence on power release inside of neighbouring fuel pins in a WWER-440 type core

    International Nuclear Information System (INIS)

    Mikus, J.

    2006-01-01

    As known both the WWER-440 and WWER-1000 reactors are systematically modernized to enhance their safety and economical parameters of operation. For this purpose new fuel assemblies (FAs) were designed with improved technical parameters, e.g., containing fuel pins (FPs) in which Gd 2 O 3 burnable absorber is integrated into fuel. Presence of such FPs in reactor core results in a strong depression of thermal neutrons in their positions and corresponding high gradients in neighbouring FPs. Consequently, similar situation in neighbouring FPs can be expected as for both the power release and temperature gradients. The purpose of this work consists in investigation of the gadolinium FP influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of the neighbouring FPs that could result in static loads with some consequences, e.g., a contribution to FP/FA bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, needed power release values inside of investigated FPs, can be estimated. For this purpose, experimental results from light water, zero-power research reactor LR-0 obtained by measurements in a WWER-440 type core with 19 FAs at zero boron concentration and containing some FPs with gadolinium (Gd FPs) were utilized. Application of the proposed evaluation method is demonstrated on investigated FPs neighbouring a Gd FP by means of the: relative azimuthal power distribution estimation inside of investigated FPs on their fuel pellet surface in horizontal plane

  20. Experience of air transport of nuclear fuel material as type A package

    International Nuclear Information System (INIS)

    Kawasaki, Masashi; Kageyama, Tomio; Suzuki, Toru

    2004-01-01

    Special law on nuclear disaster countermeasures (hereafter called as to nuclear disaster countermeasures low) that is domestic law for dealing with measures for nuclear disaster, was enforced in June, 2000. Therefore, nuclear enterprise was obliged to report accidents as required by nuclear disaster countermeasures law, besides meeting the technical requirement of existent transport regulation. For overseas procurement of plutonium reference materials that are needed for material accountability, A Type package must be transported by air. Therefore, concept of air transport of nuclear fuel materials according to the nuclear disaster countermeasures law was discussed, and the manual including measures against accident in air transport was prepared for the oversea procurement. In this presentation, the concept of air transport of A Type package containing nuclear fuel materials according to the nuclear disaster countermeasures law, and the experience of a transportation of plutonium solution from France are shown. (author)

  1. Improvement of the vibration of the test fuel(Type-B) with a guide tube under operational condition

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seung; Yim, Jeong Sik; Lim, I. C. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The Type-B test fuel for the Hanaro has a flexible guide tube on top of the fuel to lead and guide the instrumentation wires. Depending on the flow condition in the reactor, the fuel is susceptible to vibration. During the test operation of the fuel, a fairly large amplitude vibration was observed and the possibility of flow tube contact with adjacent flow tubes, due to the excessive vibration of the fuel, and consequent wear or defect of the flow tubes were raised. Thus, to know the vibration characteristics as well as whether the flow tube contact each other, analyses of the Type-B fuel the dummy fuel were performed by BEVIRA and ANSYS. Besides the analyses, vibration tests using the dummy fuel in air and with Type-B fuel in the core at zero power under operational flow condition were executed. The results from the analyses were compared with those from tests to validate the analyses. From the deflection test of the dummy fuel in air to get the maximum displacement of the flow tube at the top, the flow tube were found to contact each other. For the prevention of the contact of the flow tubes caused by the excessive vibration of the guide tube, an additional support to the guide tube was proposed. With the additional support, analysis and in core vibration test under operational flow condition were conducted and there found to be no excessive vibration any more. 6 refs., 16 figs., 6 tabs. (Author)

  2. Effects of vehicle type and fuel quality on real world toxic emissions from diesel vehicles

    Science.gov (United States)

    Nelson, Peter F.; Tibbett, Anne R.; Day, Stuart J.

    Diesel vehicles are an important source of emissions of air pollutants, particularly oxides of nitrogen (NO x), particulate matter (PM), and toxic compounds with potential health impacts including volatile organic compounds (VOCs) such as benzene and aldehydes, and polycyclic aromatic hydrocarbons (PAHs). Current developments in engine design and fuel quality are expected to reduce these emissions in the future, but many vehicles exceed 10 years of age and may make a major contribution to urban pollutant concentrations and related health impacts for many years. In this study, emissions of a range of toxic compounds are reported using in-service vehicles which were tested using urban driving cycles developed for Australian conditions. Twelve vehicles were chosen from six vehicle weight classes and, in addition, two of these vehicles were driven through the urban drive cycle using a range of diesel fuel formulations. The fuels ranged in sulphur content from 24 to 1700 ppm, and in total aromatics from 7.7 to 33 mass%. Effects of vehicle type and fuel composition on emissions are reported. The results show that emissions of these toxic species were broadly comparable to those observed in previous dynamometer and tunnel studies. Emissions of VOCs and smaller PAHs such as naphthalene, which are derived largely from the combustion process, appear to be related, and show relatively little variability when compared with the variability in emissions of aldehydes and larger PAHs. In particular, aldehyde emissions are highly variable and may be related to engine operating conditions. Fuels of lower sulphur and aromatic content did not have a significant influence on emissions of VOCs and aldehydes, but tended to result in lower emissions of PAHs. The toxicity of vehicle exhaust, as determined by inhalation risk and toxic equivalency factor (TEF)-weighted PAH emissions, was reduced with fuels of lower aromatic content.

  3. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Lily Suparlina; Tukiran Surbakti

    2014-01-01

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  4. A novel reactor type for autothermal reforming of diesel fuel and kerosene

    International Nuclear Information System (INIS)

    Pasel, Joachim; Samsun, Remzi Can; Tschauder, Andreas; Peters, Ralf; Stolten, Detlef

    2015-01-01

    Highlights: • Development and experimental evaluation of Juelich’s novel ATR reactor type. • Constructive integration of steam generation chamber and nozzle for water injection. • Internal steam generator modified to reduce pressure drop to approx. a thirtieth. • Novel concept for ATR heat management proven to be suitable for fuel cell systems. • Reaction conditions during shut-down and start-up optimized to reduce byproducts. - Abstract: This paper describes the development and experimental evaluation of Juelich’s novel reactor type ATR AH2 for autothermal reforming of diesel fuel and kerosene. ATR AH2 overcomes the disadvantages of Juelich’s former reactor generations from the perspective of the fuel cell system by constructively integrating an additional pressure swirl nozzle for the injection of cold water and a steam generation chamber. As a consequence, ATR AH2 eliminates the need for external process configurations for steam supply. Additionally, the internal steam generator has been modified by increasing its cross-sectional area and by decreasing its length. This measure reduces the pressure drop of the steam generator from approx. 500 mbar to roughly a thirtieth. The experimental evaluation of ATR AH2 at steady state revealed that the novel concept for heat management applied in ATR AH2 is suitable for fuel cell systems at any reformer load point between 20% and 120% when the mass fractions of cold water to the newly integrated nozzle are set to values between 40% and 50%. The experimental evaluation of ATR AH2 during start-up and shut-down showed that slight modifications of the reaction conditions during these transient phases greatly reduced the concentrations of ethene, ethane, propene and benzene in the reformate. From the fuel cell system perspective, these improvements provide a very beneficial contribution to longer stabilities for the catalysts and adsorption materials

  5. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  6. Bilateral Vestibular Dysfunction Associated With Chronic Exposure to Military Jet Propellant Type-Eight Jet Fuel

    Directory of Open Access Journals (Sweden)

    Terry D. Fife

    2018-05-01

    Full Text Available We describe three patients diagnosed with bilateral vestibular dysfunction associated with the jet propellant type-eight (JP-8 fuel exposure. Chronic exposure to aromatic and aliphatic hydrocarbons, which are the main constituents of JP-8 military aircraft jet fuel, occurred over 3–5 years’ duration while working on or near the flight line. Exposure to toxic hydrocarbons was substantiated by the presence of JP-8 metabolite n-hexane in the blood of one of the cases. The presenting symptoms were dizziness, headache, fatigue, and imbalance. Rotational chair testing confirmed bilateral vestibular dysfunction in all the three patients. Vestibular function improved over time once the exposure was removed. Bilateral vestibular dysfunction has been associated with hydrocarbon exposure in humans, but only recently has emphasis been placed specifically on the detrimental effects of JP-8 jet fuel and its numerous hydrocarbon constituents. Data are limited on the mechanism of JP-8-induced vestibular dysfunction or ototoxicity. Early recognition of JP-8 toxicity risk, cessation of exposure, and customized vestibular therapy offer the best chance for improved balance. Bilateral vestibular impairment is under-recognized in those chronically exposed to all forms of jet fuel.

  7. Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

    International Nuclear Information System (INIS)

    Umbehaun, Pedro Ernesto

    2000-01-01

    This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

  8. Utilization of the experimental reactor Osiris for the study and the development of fuels of the fast neutron reactor type

    International Nuclear Information System (INIS)

    Marcon, M.; Faugere, J.L.; Genthon, J.P.; Maillot, R.

    1977-01-01

    Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O 2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation [fr

  9. The influence of weather and fuel type on the fuel composition of the area burned by forest fires in Ontario, 1996-2006.

    Science.gov (United States)

    Podur, Justin J; Martell, David L

    2009-07-01

    Forest fires are influenced by weather, fuels, and topography, but the relative influence of these factors may vary in different forest types. Compositional analysis can be used to assess the relative importance of fuels and weather in the boreal forest. Do forest or wild land fires burn more flammable fuels preferentially or, because most large fires burn in extreme weather conditions, do fires burn fuels in the proportions they are available despite differences in flammability? In the Canadian boreal forest, aspen (Populus tremuloides) has been found to burn in less than the proportion in which it is available. We used the province of Ontario's Provincial Fuels Database and fire records provided by the Ontario Ministry of Natural Resources to compare the fuel composition of area burned by 594 large (>40 ha) fires that occurred in Ontario's boreal forest region, a study area some 430,000 km2 in size, between 1996 and 2006 with the fuel composition of the neighborhoods around the fires. We found that, over the range of fire weather conditions in which large fires burned and in a study area with 8% aspen, fires burn fuels in the proportions that they are available, results which are consistent with the dominance of weather in controlling large fires.

  10. Management of the acceptance process of RTR aluminide type spent fuel

    International Nuclear Information System (INIS)

    Auziere, P.; Thomasson, J.

    2002-01-01

    A wide range of Research Test Reactor aluminide type spent fuel is already received for treatment conditioning at the La Hague reprocessing complex. Such a diversity calls for an utmost attention to be paid to all safety-related systems and technical aspects, to all regulatory and administrative constraints. Despite of such multiple data inputs and rigid constraints, a close cooperation between the Research Reactor operator and COGEMA enables to reach adequate and cost effective solutions also relevant to spent fuel having had an uneven history. The acceptance process is primarily based on the client descriptive data and status declaration issued by the Research Reactor (RR) operator under QA. This acceptance process is a key step, to be keenly scheduled as it is directly interactive with the RR evacuation plans and the La Hague industrial plant program. It is also governed by the reviews conducted by the French Safety Authority and generally translated into operational authorisations. Concerned by maintaining high safety standards, reliable and proven operational levels of its nuclear services performed in the La Hague facilities COGEMA includes, all through this acceptance process, the operating, regulatory and administrative requirements. This paper sets forth an overview of the approach implemented in the COGEMA organisation for the management of the acceptance process of RTR aluminide type spent fuel. (author)

  11. Development of a Direct Methanol Fuel Cell with Lightweight Disc Type Current Collectors

    Directory of Open Access Journals (Sweden)

    Yean-Der Kuan

    2014-05-01

    Full Text Available The direct methanol fuel cell (DMFC adopts methanol solution as a fuel suitable for low power portable applications. A miniature, lightweight, passive air-breathing design is therefore desired. This paper presents a novel planar disc-type DMFC with multiple cells containing a novel developed lightweight current collector at both the anode and cathode sides. The present lightweight current collector adopts FR4 Glass/Epoxy as the substrate with the current collecting areas located at the corresponding membrane electrolyte assembly (MEA areas. The current collecting areas are fabricated by sequentially coating a corrosion resistant layer and electrical conduction layer via the thermal evaporation technique. The anode current collector has carved flow channels for fuel transport and production. The cathode current collector has drilled holes for passive air breathing. In order to ensure feasibility in the present concept a 3-cell prototype DMFC module with lightweight disc type current collectors is designed and constructed. Experiments were conducted to measure the cell performance. The results show that the highest cell power output is 54.88 mW·cm−2 and successfully demonstrate the feasibility of this novel design.

  12. Installation, maintenance and operating manual for the Lucas-type fuel injection system of the 3 B rotary engine

    Science.gov (United States)

    1985-01-01

    The installation procedure, maintenance, adjustment and operation of a Lucas type fuel injection system for 13B rotary racing engine is outlined. Components of the fuel injection system and installation procedure and notes are described. Maintenance, adjustment, and operation are discussed.

  13. Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

    International Nuclear Information System (INIS)

    Setiyanto; Tukiran Surbakti

    2016-01-01

    In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

  14. Method of determining the composition of fuels for FBR type reactors

    International Nuclear Information System (INIS)

    Tsutsumi, Kiyoshi.

    1981-01-01

    Purpose: To improve the core safety of FBR type reactors by determining the composition of fuels composed of oxide mixture of plutonium and uranium, using a relation between specific plutonium seed and plutonium enrichment degree. Method: Relation is determined between the ratio of a specific plutonium seed for constituting plutonium oxide, for example 239 U ratio and a plutonium enrichment degree required for setting the assembly power to a constant level. The ratio of 239 U is plutonium having a given isotopic ratio is also determined. The accuracy of the 239 U ratio can be improved by the correction using the density coefficient. Then, the plutonium enrichment degree is determined using the relation determined as above based on the thus determined 239 U ratio. The composition of the fuel using oxide mixture of plutonium and uranium is determined by utilizing the thus obtained plutonium enrichment degree. (Moriyama, K.)

  15. Emissions deterioration for three alternative fuel vehicle types: Natural gas, ethanol, and methanol vehicles

    International Nuclear Information System (INIS)

    Winebrake, J.J.; Deaton, M.L.

    1997-01-01

    Although there have been several studies examining emissions from in-use alternative fuel vehicles (AFVs), little is known about the deterioration of these emissions over vehicle lifetimes and how this deterioration compares with deterioration from conventional vehicles (CVs). This paper analyzes emissions data from 70 AFVs and 70 CVs operating in the federal government fleet to determine whether AFV emissions deterioration differs significantly from CV emissions deterioration. The authors conduct the analysis on three alternative fuel types (natural gas, methanol, and ethanol) and on five pollutants (carbon monoxide, carbon dioxide, total hydrocarbons, non-methane hydrocarbons, and nitrogen oxides). They find that for most cases they studied, deterioration differences are not statistically significant; however, several exceptions suggest that air quality planners and regulators must further analyze AFV emissions deterioration in order to properly include these technologies into broader air quality management schemes

  16. Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1991-10-01

    Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

  17. The relevance of the IFPE Database to the modelling of WWER-type fuel behaviour

    International Nuclear Information System (INIS)

    Killeen, J.; Sartori, E.

    2006-01-01

    The aim of the International Fuel Performance Experimental Database (IFPE Database) is to provide, in the public domain, a comprehensive and well-qualified database on zircaloy-clad UO 2 fuel for model development and code validation. The data encompass both normal and off-normal operation and include prototypic commercial irradiations as well as experiments performed in Material Testing Reactors. To date, the Database contains over 800 individual cases, providing data on fuel centreline temperatures, dimensional changes and FGR either from in-pile pressure measurements or PIE techniques, including puncturing, Electron Probe Micro Analysis (EPMA) and X-ray Fluorescence (XRF) measurements. This work in assembling and disseminating the Database is carried out in close co-operation and co-ordination between OECD/NEA and the IAEA. The majority of data sets are dedicated to fuel behaviour under LWR irradiation, and every effort has been made to obtain data representative of BWR, PWR and WWER conditions. In each case, the data set contains information on the pre-characterisation of the fuel, cladding and fuel rod geometry, the irradiation history presented in as much detail as the source documents allow, and finally any in-pile or PIE measurements that were made. The purpose of this paper is to highlight data that are relevant specifically to WWER application. To this end, the NEA and IAEA have been successful in obtaining appropriate data for both WWER-440 and WWER-1000-type reactors. These are: 1) Twelve (12) rods from the Finnish-Russian co-operative SOFIT programme; 2) Kola-3 WWER-440 irradiation; 3) MIR ramp tests on Kola-3 rods; 4) Zaporozskaya WWER-1000 irradiation; 5) Novovoronezh WWER-1000 irradiation. Before reviewing these data sets and their usefulness, the paper touches briefly on recent, more novel additions to the Database and on progress made in the use of the Database for the current IAEA FUMEX II Project. Finally, the paper describes the Computer

  18. Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence

    International Nuclear Information System (INIS)

    Silva, Clayton Pereira da

    2012-01-01

    In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U 3 O 8 and U 3 Si 2 later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U 3 Si 2 , meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous chemical treatments (dissolving

  19. Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

    International Nuclear Information System (INIS)

    Georgevich, V.; Taleyarkham, R.P.; Navarro-Valenti, S.; Kim, S.H.

    1995-01-01

    A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization

  20. Physics concept on the constellation type fissile fuels and its application to the prospective Th-232U Reactor

    International Nuclear Information System (INIS)

    Zhang, Jiahua

    1994-01-01

    In contrast with the conventional nuclear reactor which usually fuelled with on single fissile nuclide, a constellation type fissile fuels reactor consists of a parent nuclide such as 232 Th or 238 U and its whole family of neutron generated daughter nuclides. All of them are regarded as fissile fuels but of quite different fission ability. The concentration of each daughter nuclide is determined by its saturate concentration ratio with the parent nuclide. In such fuel system, the whole fuel consumed by neutron reaction almost completely results in fission products. In this article, some properties of such fuel system, determination of the saturate concentration of each daughter nuclide and applicability to Th- 233 U fueled reactor will be discussed. 3 refs., 1 tab., 2 figs

  1. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1992-11-01

    This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements

  2. Preparation of U3O8 powder for MTR type fuel from ammonium uranyl carbonate

    International Nuclear Information System (INIS)

    Marcondes, G.H.; Riella, H.G.

    1990-08-01

    In this paper it is described the research done at IPEN-CNEN/SP on the preparation of U 3 O 8 powder from calcination of the AUC, with appropriate characteristics to be used as dispersoid for MTR type fuel. The calcination in air of the AUC leads a U 3 O 8 powder that is further processed to obtain a powder with density and particle size as especifications. The important process parameters are here discussed with the variation AUC calcination temperature and sintering time of the U 3 O 8 powder. (author) [pt

  3. Burn-up calculations for a thorium HTR with one and with two types of fuel particle

    Energy Technology Data Exchange (ETDEWEB)

    Griggs, C. F.

    1975-06-15

    Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.

  4. On the possibility of reprocessing of fuel elements of dispersion type with copper matrix by pyrochemical methods

    International Nuclear Information System (INIS)

    Vasin, B.D.; Ivanov, V.A.; Shchetinskij, A.V.; Vavilov, S.K.; Savochkin, Yu.P.; Bychkov, A.V.; Kormilitsyn, M.V.

    2005-01-01

    A consideration is given to pyrochemical processes suitable for separation of uranium dioxide from structural materials when reprocessing cermet type fuel elements. The estimation of the possibility to apply liquid antimony and bismuth, potassium and copper chlorides melts is made. The specimens compacted of copper and uranium dioxide powders in a stainless steel can are used as simulators of fuel element sections. It is concluded that the dissolution of structural materials in molten salts at the stage of uranium dioxide concentration is the process of choice for reprocessing of dispersion type fuel elements [ru

  5. Utilization of radiographic and ultrasonic testing for an evaluation of plate type fuel elements during manufacturing stages

    International Nuclear Information System (INIS)

    Brito, Mucio Jose Drummond de; Silva Junior, Silverio Ferreira da; Messias, Jose Marcos; Braga, Daniel Martins; Paula, Joao Bosco de

    2005-01-01

    Structural discontinuities can be introduced in the plate type fuel elements during the manufacturing stages due to mechanical processing conditions. The use of nondestructive testing methods to monitoring the fuel elements during the manufacturing stages presents a significant importance, contributing for manufacturing process improvement and cost reducing. This paper describes a procedure to be used detection and evaluation of structural discontinuities in plate type fuel elements during the manufacturing stages using the ultrasonic testing method and the radiographic testing method. The main results obtained are presented and discussed. (author)

  6. A Study on BC Emission from Vehicles using Different Types of Fuel

    Science.gov (United States)

    Kim, K.; Son, J.; Kim, J.; Kim, S.; Park, G.; Sung, K.; Kim, I.; Chung, T.; Park, T.; Kang, S.; Ban, J.; Kim, J.; Hong, Y. D.; Woo, J. H.; Lee, T.

    2017-12-01

    Black carbon (BC) is an anthropogenic aerosol from fossil fuels, and biomass burning. It absorbs solar radiation, and heats the atmosphere leading 0.4W m-2 radiative forcing. BC is a particle that can cause serious effects on human body as well. Toxicological studies of black carbon suggests that BC may be an important carrier of toxic chemicals to human body. The recent researches show that one of the main precursor of BC is vehicle emission, but the inventory of BC emission rate from vehicle is inadequate in South Korea. This study tries to find differences of BC emission from different sizes of vehicles using different types of fuels. Fuels used in vehicles are gasoline, liquefied petroleum gas (LPG), and diesel. BC was directly measured from the tail pipe of vehicles using Aethalometer (AE33, Magee Scientific Corporation). This study was conducted in Transport Pollutant Research Center, National Institute of Environmental Research, South Korea. Measurement was progressed with the five different test modes of speeds. Speed modes includes 4.7, 17.3, 34.1, 65.4, and 97.3 km h-1. Emission rate of BC was high in the slowest speed mode, and showed decrease with increase of the speed of vehicles. Gasoline vehicles had the relatively higher emission rate of BC than the LPG vehicle, while the emission rate of BC for Diesel with DPF (Diesel Particle Filter) was observed to be the lowest.

  7. Performance enhancement of a spark ignition engine fed by different fuel types

    International Nuclear Information System (INIS)

    Hedfi, Hachem; Jbara, Abdessalem; Jedli, Hedi; Slimi, Khalifa; Stoppato, Anna

    2016-01-01

    Highlights: • Biogas mixed with hydrogen is checked for a spark ignition engine. • An engine fed by biogas, hydrogen, natural gas or liquid petroleum gas is studied. • Efficiency is optimized with respect to consumption and exhaust gas recirculation. • Combustion reaction progress is characterized in real time. - Abstract: A numerical model based on thermodynamic and kinetic analyses has been established in order to evaluate biogas, hydrogen, natural gas or liquid petroleum gas as fuels in a spark ignition engine. For each fuel type, consumption as well as efficiency have been compared to gasoline in order to generate the same engine work (in the range of 0.28–0.43 W h/cycle). It was found that the spark ignition engine can be fed by an equimolar mixture of biogas and hydrogen. Moreover, thermal efficiency has been enhanced with respect to fuel consumption and exhaust gas recirculation (EGR). It was shown that an equimolar mixture between biogas and hydrogen increases the ITE by around 2.2% and decreases the mass consumption by less than 0.01 g/cycle. In addition, the combustion reaction progresses as well as CO and CO_2 emissions have been characterized in real time.

  8. Vibration characteristics of a PWR fuel rod supported by optimized H type spacer grids

    International Nuclear Information System (INIS)

    Choi, M. H.; Kang, H. S.; Yoon, K. H.; Kim, H. K.; Song, K. N.

    2002-01-01

    The spacer grids are one of the main structural components in the fuel assembly, which supports and protects the fuel rods from the external loads by seismic and coolant flow. In this study, a modal test and a FE vibration analysis using ABAQUS are performed on a PWR dummy fuel rod of 3.847 m which is continuously supported by eight Optimized H type spacer grids. The experimental results agree with previous works that the natural frequencies decrease, while the amplitudes increase, with the increase of the excitation force. The force levels showing the maximum displacement of 0.2 mm are in the range from 0.2 N to 0.3 N, and at the same force range the fundamental frequencies are measured around 42.0 Hz, at which the relatively big displacements are observed at the 7th span. The results from the modal tests and the FE analyses are compared by both Modal Assurance Criteria (MAC) values and mode shapes. The MAC values at 2nd, 4th, and 7th mode are below 50%. It is believed that the reason of the low MACs at those modes is that the vibration amplitudes of the modes are more distorted by the excitation force than those of the other modes

  9. Fabrication of AA6061-T6 Plate Type Fuel Assembly Using Electron Beam Welding Process

    International Nuclear Information System (INIS)

    Kim, Soosung; Seo, Kyoungseok; Lee, Donbae; Park, Jongman; Lee, Yoonsang; Lee, Chongtak

    2014-01-01

    AA6061-T6 aluminum alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW. However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the shrinkage measurement and weld inspection using computed tomography. This study was carried out to determine the suitable welding parameters and to evaluate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory electron beam welding process of the full-sized sample was being developed. Based on this fundamental study, fabrication of the plate-type fuel assembly will be provided for the future Ki-Jang research reactor project

  10. Gerenciamento de Projetos Segundo o Guia PMBOK: Desafios para os Gestores

    Directory of Open Access Journals (Sweden)

    David Ferreira Bomfin

    2013-01-01

    Full Text Available Analisamos, dentre as dificuldades e facilidades mapeadas na literatura, quais e por que as mesmas estão presentes no processo de gerenciamento de projetos com base no guia PMBOK. Verificamos quais os possíveis argumentos que os gestores vivenciam no gerenciamento de projetos no que diz respeito aos sucessos e fracassos apresentados. O estudo se concentrou em profissionais com certificação PMP por apresentarem o conhecimento em Gerenciamento de Projetos com base no guia PMBOK e, por ser uma área de crescimento onde as empresas buscam profissionais com melhores práticas e vantagem competitiva. Valendo-se do questionário estruturado como instrumento de coleta de dados e analisando os resultados, os principais pontos encontrados dizem respeito ao tempo em que os gerentes lidam com gerenciamento de projetos, as dificuldades encontradas pelos gerentes em sua prática profissional, a área do PMBOK com maior dificuldade de ser gerenciada, os itens que geram sucessos e fracassos nos projetos. A entrevista semi-estruturada evidenciou que o gerenciamento de projetos apresenta algumas dificuldades presentes na prática profissional dos gerentes. Sugerimos investigar melhor o impacto que a certificação PMP proporciona aos gerentes de projetos e se a relação maturidade organizacional e experiência em gerenciamento de projetos é fator de sucesso para o projeto.DOI:10.5585/gep.v3i3.78

  11. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  12. Hydrodeoxygenation of oxidized distilled bio-oil for the production of gasoline fuel type

    International Nuclear Information System (INIS)

    Luo, Yan; Guda, Vamshi Krishna; Hassan, El Barbary; Steele, Philip H.; Mitchell, Brian; Yu, Fei

    2016-01-01

    Highlights: • Oxidation had more influence on the yield of total hydrocarbons than distillation. • The highest total hydrocarbon yield was obtained from oxidized distilled bio-oil. • The 2nd-stage hydrocarbons were in the range of gasoline fuel boiling points. • The main products for upgrading of oxidized bio-oil were aliphatic hydrocarbons. • The main products for upgrading of non-oxidized bio-oil were aromatic hydrocarbons. - Abstract: Distilled and oxidized distilled bio-oils were subjected to 1st-stage mild hydrodeoxygenation and 2nd-stage full hydrodeoxygenation using nickel/silica–alumina catalyst as a means to enhance hydrocarbon yield. Raw bio-oil was treated for hydrodeoxygenation as a control to which to compare study treatments. Following two-stage hydrodeoxygenation, four types of hydrocarbons were mainly comprised of gasoline and had water contents, oxygen contents and total acid numbers of nearly zero and higher heating values of 44–45 MJ/kg. Total hydrocarbon yields for raw bio-oil, oxidized raw bio-oil, distilled bio-oil and oxidized distilled bio-oil were 11.6, 16.2, 12.9 and 20.5 wt.%, respectively. The results indicated that oxidation had the most influence on increasing the yield of gasoline fuel type followed by distillation. Gas chromatography/mass spectrometry characterization showed that 66.0–76.6% of aliphatic hydrocarbons and 19.5–31.6% of aromatic hydrocarbons were the main products for oxidized bio-oils while 35.5–38.7% of aliphatic hydrocarbons and 58.2–63.1% of aromatic hydrocarbons were the main products for non-oxidized bio-oils. Both aliphatic and aromatic hydrocarbons are important components for liquid transportation fuels and chemical products.

  13. Effects of spent fuel types on offsite consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-01-01

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor (χ/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only 134 Cs, 137 Cs- 137m Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents

  14. Evaluation of Electron Beam Welding Performance of AA6061-T6 Plate-type Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Soo-Sung; Seo, Kyoung-Seok; Lee, Don-Bae; Park, Jong-Man; Lee, Yoon-Sang; Lee, Chong-Tak

    2014-01-01

    As one of the most commonly used heat-treatable aluminum alloys, AA6061-T6 aluminum alloy is available in a wide range of structural materials. Typically, it is used in structural members, auto-body sheet and many other applications. Generally, this alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW(Electron Beam Welding). However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the plate-type nuclear fuel fabrication and assembly, a fundamental electron beam welding experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the suitable welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the plate-type fuel assembly has been also studied by the weld penetrations of side plate to end fitting and fixing bar and weld inspections using computed tomography

  15. Control console conceptual design for sheet type fuels of Triga Mark-II reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Kurnia Wibowo; Anang Susanto

    2016-01-01

    The control console conceptual design for sheet type fuel of TRIGA Mark-II reactor has been made. The control console conceptual design was made with refer study result of instrument and control system which is used in BATAN'S reactor i.e TRIGA-2000 Bandung, TRIGA Yogyakarta and MPR-30 Serpong. The control console conceptual design was made by using AutoCad software. The control console conceptual design reactor for sheet type fuel of TRIGA Mark-II reactor consist of 5 segments that is 3 segments for placing the computer monitors, 1 segment for placing bargraph displays and recorders and 1 segment for placing panel meters. There are the door on front and back position at each segment for enter and out devices in the console. The control console conceptual design is also equipped by the table along in front of console for placing reactor panel control and for writing, 3 drawers for 3 keyboards. The dimension of console will refer control room size and the components will be placed on console which will be detailed in detail design if this conceptual design has been approved. (author)

  16. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  17. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    International Nuclear Information System (INIS)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S.

    2013-01-01

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  18. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  19. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Jiang Yijie; Wang Qiming; Cui Yi; Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Ding Shurong, E-mail: dsr1971@163.com [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

    2011-06-15

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  20. Taguchi Method for Investigating the Performance Parameters and Exergy of a Diesel Engine Using Four Types of Diesel Fuels

    Directory of Open Access Journals (Sweden)

    Dara K. Khidir

    2016-05-01

    Full Text Available The effects of changes in engine operating parameters, i.e., engine speed, throttle and water temperature, for four types of diesel fuel (A, B, C and D of different specific gravities, as supplied from local market and refineries, were studied and simultaneously optimized. The experiment design was based on Taguchi’s “L' 16” orthogonal table, and the engine was put to test at different engine speeds, throttling opening percentages and water temperatures, using different fuels. The data were analyzed using S/N (signal to noise ratio for each factor. The obtained results show that the optimum operating conditions for minimum BSFC (brake specific fuel consumption are achieved when the engine speed is 2500 rpm, the throttle is placed at 75% of full throttling, the water temperature is 80 oC and the engine is using fuel type D. Also, results of S/N ratio reveal that the throttle has significant influence on brake thermal and exergic efficiencies. Water temperature is the second most effective factor and then comes the influence of engine speed. The least effective factor among the studied parameters for the types of fuel considered in this experiment is the fuel type.

  1. Metal membrane-type 25-kW methanol fuel processor for fuel-cell hybrid vehicle

    Science.gov (United States)

    Han, Jaesung; Lee, Seok-Min; Chang, Hyuksang

    A 25-kW on-board methanol fuel processor has been developed. It consists of a methanol steam reformer, which converts methanol to hydrogen-rich gas mixture, and two metal membrane modules, which clean-up the gas mixture to high-purity hydrogen. It produces hydrogen at rates up to 25 N m 3/h and the purity of the product hydrogen is over 99.9995% with a CO content of less than 1 ppm. In this fuel processor, the operating condition of the reformer and the metal membrane modules is nearly the same, so that operation is simple and the overall system construction is compact by eliminating the extensive temperature control of the intermediate gas streams. The recovery of hydrogen in the metal membrane units is maintained at 70-75% by the control of the pressure in the system, and the remaining 25-30% hydrogen is recycled to a catalytic combustion zone to supply heat for the methanol steam-reforming reaction. The thermal efficiency of the fuel processor is about 75% and the inlet air pressure is as low as 4 psi. The fuel processor is currently being integrated with 25-kW polymer electrolyte membrane fuel-cell (PEMFC) stack developed by the Hyundai Motor Company. The stack exhibits the same performance as those with pure hydrogen, which proves that the maximum power output as well as the minimum stack degradation is possible with this fuel processor. This fuel-cell 'engine' is to be installed in a hybrid passenger vehicle for road testing.

  2. Verification test of advanced LWR fuel components of Westinghouse type nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho

    2004-08-01

    The purpose of this project is to independently conduct the performance test of the spacer grids and the cladding material of the 16x16 and 17x17 advanced fuels for Westinghouse type plants, and to improve the relevant test technology. Major works and results of the present research are as follows. 1. The design and structural features of the spacer grids were investigated, especially the finally determined I-spring was thoroughly analyzed in the point of the mechanical damage and characteristic. 2. As for the mechanical tests of the space grids, the characterization, the impact and the fretting wear tests were carried out. The block as well as the in-grid tests were conducted for the spring/dimple characterization, from which a simple method was developed that simulated the boundary conditions of the assembled grid straps. The impact tester was modified and improved to accommodate a full size grid assembly. The impact result showed that the grid assembly fulfilled the design criteria. As for the fretting wear tests, a sliding test under the room temperature air/water, a sliding/impact test under the room temperature air and a sliding/impact tests under the high temperature and pressure environments were carried out. To this end, a high temperature and pressure fretting wear tester was newly developed. The wear characteristic and the resistibility of the advanced grid spring/dimple were analyzed in detail. The test results were verified through comparing those with the test results by the Westinghouse company. 3. The properties and performance of the newly adopted material for the cladding, Low Sn Zirlo was investigated by a room and high temperature tensile tests and a corrosion tests under the environments of 360 .deg. C water, 400 steam and 360 .deg. C 70ppm LiOH. Through the present project, all the test equipment and technologies for the fuel components were procured, which will be used for future domestic development of a new fuel

  3. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1993-01-01

    Three furnace heating tests were conducted with irradiated, HT9-clad and U-19wt%Pu-10wt%Zr-alloy, EBR-II Mk-V-type fuel elements to evaluate the behavior that could be expected during a loss-of-flow event in the reactor. In general, very significant safety margins for cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results are presented, as are discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction that were found in these tests. (orig.)

  4. Application of fire-retardant treatment to the wood in Type A unirradiated nuclear fuel outer containers

    International Nuclear Information System (INIS)

    Whitlow, J.D.; Luna, R.E.

    1992-01-01

    Packagings for transporting unirradiated nuclear fuel assemblies in the United States are commonly constructed as rectangular boxes consisting of a metal inner container, a wooden outer container, and cushioning material separating the two. The wood in the outer container is a potential source of fuel for fire. Use of a fire-retardant treatment on the wood may reduce or eliminate the damage to nuclear fuel assemblies in some types of accidents involving fire. The applicability of using fire-retardant treatments on the wood of outer containers is addressed. An approximate cost-benefit analysis to determine if fire-retardant treatments are economically justified is presented. (Author)

  5. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    International Nuclear Information System (INIS)

    Uriarte, A.; Ramos, L.; Estrada, J.; del Val, J. L.

    1962-01-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO 2 F 2 solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs

  6. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

  7. Postirradiation Examination Of U3O8-AL Plate Type Dispersion Fuel Element

    International Nuclear Information System (INIS)

    Nasution-Hasbullah; Sugondo; Amin, D.L.; Siti-Amini

    1996-01-01

    Postirradiation examination of plate type spent fuel element RIE-01 has been carried out in order to observer its physical changes and performance under irradiation in the reactor. The irradiation has been time more than two years with a declared burnup of 51.04 %. The examination included visual and dimensional measurement, measurement of burn-up distribution, wipe test and metallographic analysis. The results showed that all fuel plates retained their integrity. The colour changes were occurred on most of the plates significant suggesting that it was generated from the oxide layer formation. From gamma-scanning examination it could be deducted that the highest burn-up distribution of the plate was at position of 30 cm from the bottom. A more homogeneous distribution was found in the middle plate of the bundle. The increased plate thickness, as revealed by dimensional measurements as in agreement with the burn-up distribution pattern. Despite the changes observed in could be concluded that all changes occurred were still within the allowable limits and therefore it can recommended that an increase of the burn-up level above 51,04 % is still quite possible

  8. Pressure drop calculation in a fuel element of a pool type reactor

    International Nuclear Information System (INIS)

    Lassance, Victor; Oliveira, Andre F.; Moreira, Maria de L.

    2013-01-01

    Even with the advances of hardware in computer sciences, sometimes it is necessary to simplify the simulation in order to optimize the results given the same calculation runtime. The object of this study is a thermodynamic analysis of the core of a pool type research reactor, focusing on natural circulation. Due to the high geometrical complexity of the core, the scale transfer process becomes an essential step to the thermodynamic study of the reactor. This process takes place by determining the effective equivalent properties obtained from a detailed simulation of the core and transferring them to a porous medium having a coarse mesh while preserving the overall characteristics. In this way, it will be able to obtain the quadratic resistance coefficient KQ by calculating the pressure drop inside the fuel element. To observe in detail the behavior of this flow, longitudinal and transversal cross sections will be made in different points, thereby observing the velocity and pressure distributions. The analysis will provide detailed data on the fluid flow between the fuel plates enabling the observation of possible critical points or undesired behavior. The whole analysis was made by using the commercial code ANSYS CFX ver. 12.1. This is study will provide data, as a first step to enable future simulations which will consider the entire reactor. (author)

  9. Results of experiments with flare type igniters on diesel fuel and crude oil emulsions

    International Nuclear Information System (INIS)

    Moffat, C.; Hankins, P.

    1997-01-01

    Development of a hand-deployable igniter that could ignite contained diesel fuel and crude oil emulsions on water was described. The igniter was developed as part of the U.S. Navy Supervisor of Salvage (SUPSALV) In-Situ Burn (ISB) system. It is a manually operated, electrically fired, high temperature flare type igniter. It is 41 cm long, 10 cm in diameter, weighs 1.5 kg, and is packaged and shipped with the ISB system. The chemical and mineral composition of the flair allows for a three minute burn of up to 1370 degrees C (2500 degrees F) at the center. The flare is most effective when used in conjunction with a shroud of sorbent material which traps and holds oil around the burning flare aiding the ignition process by increasing the initial propagation area. In small-scale tank experiments the flare ignited diesel fuel in ambient temperatures of 3 degrees C, with winds of 8 to 10 m/sec. The flare also ignited 22.5 per cent water-in crude oil emulsion in 3 degrees C temperatures. 4 refs., 3 tabs

  10. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  11. Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

    1991-08-01

    According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

  12. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    Kozma, R.; Turkcan, E.; Verhoef, J.P.

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

  13. Seismic response of high temperature gas-cooled reactor core with block-type fuel, (2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1980-01-01

    For the aseismic design of a high temperature gas-cooled reactor (HTGR) with block-type fuel, it is necessary to predict the motion and force of core columns and blocks. To reveal column vibration characteristics in three-dimensional space and impact response, column vibration tests were carried out with a scale model of a one-region section (seven columns) of the HTGR core. The results are as follows: (1) the column has a soft spring characteristic based on stacked blocks connected with loose pins, (2) the column has whirling phenomena, (3) the compression spring force simulating the gas pressure has the effect of raising the column resonance frequency, and (4) the vibration behavior of the stacked block column and impact response of the surrounding columns show agreement between experiment and analysis. (author)

  14. Structural analysis on the open basket type instrumented capsule for fuel irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Sik; Kang, Y. H.; Kim, B. G.; Cho, M. S.; Sohn, J. M.; Choo, K. N.; Oh, J. M.; Shin, Y. T.; Park, S. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    To develop the open basket type instrumented capsule to be used for the irradiation test of various nuclear fuels, it is necessary to ensure the compatibility of the capsule with HANARO and the structural integrity of the capsule. The dimensions of the open basket type instrumented capsule were determined in the basis of the pressure drop criteria in OR test hole of HANARO(mass flow rate <12.7kg/s, pressure drop {delta}P>200kPa). From the buckling stability analysis for this capsule, the critical buckling load P{sub cr} was 7.5kN. The vertical impact stress of the capsule under unit impact load was evaluated by the transient analysis, and the maximum vertical impact load calculated from the impact stress and the allowable stress was 60.5kN. Under the loading of the calculated Pcr, the maximum vertical impact stress was 20.4MPa. The structural integrity of the capsule under a horizontal impact loading was also examined. The mechanical stresses occurred by the pressure difference at the inner and outer surface of cladding and by the coolant pressure at the surface of cladding were 3.1MPa and 43.3MPa, respectively. These stress values were lower than the allowable stress in each case. Therefore, it was ensured that the instrumented capsule for the irradiation test of various nuclear fuels met the criteria on the structural integrity during installing and testing the capsule in HANARO. 8 refs., 61 figs., 3 tabs. (Author)

  15. IRF3 and type I interferons fuel a fatal response to myocardial infarction.

    Science.gov (United States)

    King, Kevin R; Aguirre, Aaron D; Ye, Yu-Xiang; Sun, Yuan; Roh, Jason D; Ng, Richard P; Kohler, Rainer H; Arlauckas, Sean P; Iwamoto, Yoshiko; Savol, Andrej; Sadreyev, Ruslan I; Kelly, Mark; Fitzgibbons, Timothy P; Fitzgerald, Katherine A; Mitchison, Timothy; Libby, Peter; Nahrendorf, Matthias; Weissleder, Ralph

    2017-12-01

    Interferon regulatory factor 3 (IRF3) and type I interferons (IFNs) protect against infections and cancer, but excessive IRF3 activation and type I IFN production cause autoinflammatory conditions such as Aicardi-Goutières syndrome and STING-associated vasculopathy of infancy (SAVI). Myocardial infarction (MI) elicits inflammation, but the dominant molecular drivers of MI-associated inflammation remain unclear. Here we show that ischemic cell death and uptake of cell debris by macrophages in the heart fuel a fatal response to MI by activating IRF3 and type I IFN production. In mice, single-cell RNA-seq analysis of 4,215 leukocytes isolated from infarcted and non-infarcted hearts showed that MI provokes activation of an IRF3-interferon axis in a distinct population of interferon-inducible cells (IFNICs) that were classified as cardiac macrophages. Mice genetically deficient in cyclic GMP-AMP synthase (cGAS), its adaptor STING, IRF3, or the type I IFN receptor IFNAR exhibited impaired interferon-stimulated gene (ISG) expression and, in the case of mice deficient in IRF3 or IFNAR, improved survival after MI as compared to controls. Interruption of IRF3-dependent signaling resulted in decreased cardiac expression of inflammatory cytokines and chemokines and decreased inflammatory cell infiltration of the heart, as well as in attenuated ventricular dilation and improved cardiac function. Similarly, treatment of mice with an IFNAR-neutralizing antibody after MI ablated the interferon response and improved left ventricular dysfunction and survival. These results identify IRF3 and the type I IFN response as a potential therapeutic target for post-MI cardioprotection.

  16. The technique for determination of surface contamination by uranium on U3Si2-Al plate-type fuel elements

    International Nuclear Information System (INIS)

    Li Shulan; He Fengqi; Wang Qingheng; Han Jingquan

    1993-04-01

    The NDT method for determining the surface contamination by uranium on U 3 Si 2 -Al plate-type fuel elements, the process of standard specimen preparation and the graduation curve are described. The measurement results of U 3 Si 2 -Al plate-type fuel elements show that the alpha counting method to measure the surface contamination by uranium on fuel plate is more reliable. The UB-1 type surface contamination meter, which was recently developed, has many advantages such as high sensitivity to determine the uranium pollution, short time in measuring, convenience for operation, and the minimum detectable amount of uranium is 5 x 10 -10 g/cm 2 . The measuring device is controlled by a microcomputer. Besides data acquisition and processing, it has functions of statistics, output data on terminal or to printer and alarm. The procedures of measurement are fully automatic. All of these will meet the measuring needs in batch process

  17. Commuters' exposure to particulate matter air pollution is affected by mode of transport, fuel type, and route.

    Science.gov (United States)

    Zuurbier, Moniek; Hoek, Gerard; Oldenwening, Marieke; Lenters, Virissa; Meliefste, Kees; van den Hazel, Peter; Brunekreef, Bert

    2010-06-01

    Commuters are exposed to high concentrations of air pollutants, but little quantitative information is currently available on differences in exposure between different modes of transport, routes, and fuel types. The aim of our study was to assess differences in commuters' exposure to traffic-related air pollution related to transport mode, route, and fuel type. We measured particle number counts (PNCs) and concentrations of PM2.5 (particulate matter bus passengers, we calculated that the inhaled air pollution doses were highest for cyclists. With the exception of PM10, we found that inhaled air pollution doses were lowest for electric bus passengers. Commuters' rush hour exposures were significantly influenced by mode of transport, route, and fuel type.

  18. Guia per al càlcul simplificat de corbes de valoració

    OpenAIRE

    Núñez Burcio, Oscar; Serrano i Plana, Núria; Subirats i Vila, Xavier

    2016-01-01

    La guia descriu com calcular corbes de valoració d'acids i bases, utilitzant simplificacions, a partir de la seva concentració i de les constants de dissociació. Els casos presentats són els següents: - Valoració d’un àcid fort amb una base forta - Valoració d’una base forta amb un àcid fort - Valoració d’un àcid feble monopròtic amb una base forta - Valoració d’una base feble monopròtica amb un àcid fort - Valoració d’un àcid feble dipròtic amb una base forta - Valoració d...

  19. Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

    International Nuclear Information System (INIS)

    Edy-Sulistyono

    1996-01-01

    Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

  20. Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

    International Nuclear Information System (INIS)

    Misra, B.; Davis, J.F.

    1980-02-01

    The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio

  1. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel

    International Nuclear Information System (INIS)

    Francois, J.L.; Nunez C, A.

    2003-01-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  2. Reactor core T-H characteristics determination in case of parallel operation of different fuel assembly types

    International Nuclear Information System (INIS)

    Hermansky, J.; Petenyi, V.; Zavodsky, M.

    2009-01-01

    The WWER-440 nuclear fuel vendor permanently improve the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. Therefore it is necessary to have the skilled methodology and computing code for analyzing factors which affecting the accuracy of flow redistributed determination through reactor on flows through separate parts of reactor core in case of parallel operation different assembly types. Whereas the geometric parameters of new manufactured assemblies were changed recently, the calculated flows through the fuel parts of different type of assemblies are depended also on their real position in reactor core. Therefore the computing code CORFLO was developed in VUJE Trnava for carrying out stationary analyses of T-H characteristics of reactor core within 60 deg symmetry. The CORFLO code deals the area of the active core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is calculated. Computing code is verified and validated at this time. Paper presents the short description of computing code CORFLO with some calculated results. (Authors)

  3. Nuclear fuel burnup calculation in a Voronezh type reactor; Analiza izgaranja nuklearnog goriva u reaktoru tipa Voronjez

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M; Marinkovic, N; Kocic, A [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1977-07-01

    In order to summarize and present our abilities to perform a complex computation of the nuclear fuel burn-up, a systematic review of the available methods, algorithms and computer programmes is given in this paper. The computer programmes quoted have all been developed, modified and tested in our department, so that they can be successfully used in the analysis of nuclear power plants from both physics and economic points of view. For a commercially proven nuclear reactor - reactor of the Voronezh type - an illustrative computation of the fuel burn-up is performed. The typical results are presented and discussed. The conclusion concerns the completion of a modular scheme for the fuel burn-up calculation and the fuel cycle analysis (author)

  4. High performance liquid chromatographic hydrocarbon group-type analyses of mid-distillates employing fuel-derived fractions as standards

    Science.gov (United States)

    Seng, G. T.; Otterson, D. A.

    1983-01-01

    Two high performance liquid chromatographic (HPLC) methods have been developed for the determination of saturates, olefins and aromatics in petroleum and shale derived mid-distillate fuels. In one method the fuel to be analyzed is reacted with sulfuric acid, to remove a substantial portion of the aromatics, which provides a reacted fuel fraction for use in group type quantitation. The second involves the removal of a substantial portion of the saturates fraction from the HPLC system to permit the determination of olefin concentrations as low as 0.3 volume percent, and to improve the accuracy and precision of olefins determinations. Each method was evaluated using model compound mixtures and real fuel samples.

  5. ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

    Directory of Open Access Journals (Sweden)

    Setiyanto Setiyanto

    2016-10-01

    Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

  6. In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-06-15

    In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

  7. Imaging of Flames in Cement Kilns To Study the Influence of Different Fuel Types

    DEFF Research Database (Denmark)

    Pedersen, Morten Nedergaard; Nielsen, Mads; Clausen, Sønnik

    2017-01-01

    The cement industry aims to use an increased amount of alternative fuels to reduce production costs and CO2 emissions. In this study three cement plants firing different kinds and percentages of alternative fuel were studied. A specially developed camera setup was used to monitor the flames...... in the three cement kilns and assess the effect of alternative fuels on the flame. It was found that cofiring with solid recovered fuel (SRF) would delay the ignition point by about 2 m and lower the intensity and temperature of the kiln flame compared to a fossil fuel flame. This is related to a larger...... particle size and moisture content of the alternative fuels, which lowers the conversion rate compared to fossil fuels. The consequences can be a lower kiln temperature and cement quality. The longer conversion time may also lead to the possibility of localized reducing conditions in the cement kiln, which...

  8. Powering Profits. Profits, Investments and Fuel Type Mixes in the Dutch Power Sector

    International Nuclear Information System (INIS)

    Wilde-Ramsing, J.; Steinweg, T.

    2007-06-01

    This report addresses the Dutch power sector, identifying the major corporate players in the market, types of fuel used to generate electricity, the profits being made, and investments in both renewable and non-renewable generation capacity. For the purposes of this report, the power sector is understood to encompass production (i.e. generation) and supply of electricity. Some discussion and figures on heat and gas, which are also essential energy services, are provided, but the focus is primarily on electricity. Section 2 of the report provides an overview of the Dutch power sector, breaking the market down into production and supply. Major players, markets shares, and recent trends and developments are given for each of these activities. Sections 3 - 7 go into detail on the five major corporate players active in the Dutch power sector: ENECO, Essent, Nuon, Electrabel, and E.ON Energie. For each company, information is provided on profits and earnings, the fuel mix used to generate and supply electricity, the CO2 emissions associated with these activities, installed capacity in the Netherlands, and recent investments in renewable and non-renewable generation capacity in the Netherlands. For the Dutch companies, ENECO, Essent and Nuon, additional information on the ownership structure of the company, shareholders and dividends paid and received is given. A section on RWE (Section 8) is also included in the study because, although RWE is not currently active in generating electricity in the Netherlands, RWE Energy does currently supply electricity generated by producers in the Netherlands. In addition, RWE Power is currently planning to invest significantly in power generation capacity in the Netherlands. The final section of the report compares the companies activities in the Netherlands and draws conclusions based on the companies' respective performance

  9. Improvement of critical heat flux correlation for research reactors using plate-type fuel

    International Nuclear Information System (INIS)

    Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio

    1998-01-01

    In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as Loss of the primary coolant flow'. Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study. (author)

  10. Nuclear power plant types and the management of plutonium and minor actinides - in search of fuel cycle flexibility

    International Nuclear Information System (INIS)

    Thomas, J.B.

    2002-01-01

    Transuranics management concerns all NPP types, because of the specifications for sustainable development. Multiple recycling is mandatory. Neutronic abundance can be obtained in fast spectrum, or by adding external neutrons or (temporarily) with additional 235 U. The LWRs can control the plutonium inventory and significantly reduce the amount of transuranics transferred to the geological repository, thanks to the use of innovative nuclear fuel in a limited part of the NPP fleet. HTR adapted to transuranics burning can help. In the future, in addition to the liquid metal FBR, a strategy based on a gas cooled technological line and advanced fuel opens a second path towards fast spectra. Strategies for defining the optimal mix of reactor types in the nuclear fleet at a given time and demonstrating the fuel cycle flexibility are under study. (author)

  11. O que visitar em Paris durante a Exposição Universal de 1878: um guia turístico para geólogos

    Directory of Open Access Journals (Sweden)

    Maria Margaret Lopes

    2015-10-01

    Full Text Available Este artigo parte da tradição constituída dos guias de turismo, para comentar os guias elaborados para as Exposições Universais. Apresenta um exemplo específico desses guias de turismo científico preparado para um Congresso que se realizou no âmbito de uma Exposição e que se tornaria modelar para os próximos eventos da área. Trata-se do Guide du géologue à l’Exposition universelle de 1878 et dans les collections publiques et privées de Paris. O Guide foi organizado para orientar os geólogos estrangeiros que participaram do Primeiro Congresso Internacional de Geologia realizado durante a Exposição Universal de Paris de 1878.

  12. PLACA/DPLACA: a code to simulate the behavior of a monolithic/dispersed plate type fuel

    International Nuclear Information System (INIS)

    Denis, Alicia; Soba, Alejandro

    2005-01-01

    The PLACA code was originally built to simulate monolithic plate fuels contained in a metallic cladding, with a gap in between. The international program of high density fuels was recently oriented to the development of a plate-type fuel of a uranium rich alloy with a molybdenum content between 6 to 10 w %, without gap and with a Zircaloy cladding. To give account of these fuels, the DPLACA code was elaborated as a modification of the original code. The extension of the calculation tool to disperse fuels involves a detailed study of the properties and models (still in progress). Of special interest is the material formed by U Mo particles dispersed in an Al matrix. This material has appeared as a candidate fuel for high flux research reactors. However, the interaction layer that grows around the particles has a deleterious effect on the material performance in operation conditions and may represent a limit for its applicability. A number of recent experiments carried out on this material provide abundant information that allows testing of the numerical models. (author)

  13. Development of the Simulation Program for the In-Vessel Fuel Handling System of Double Rotating Plug Type

    International Nuclear Information System (INIS)

    Kim, S. H.; Kim, J. B.

    2011-01-01

    In-vessel fuel handling machines are the main equipment of the in-vessel fuel handling system, which can move the core assembly inside the reactor vessel along with the rotating plug during refueling. The in vessel fuel handling machines for an advanced sodium cooled fast reactor(SFR) demonstration plant are composed of a direct lift machine(DM) and a fixed arm machine(FM). These machines should be able to access all areas above the reactor core by means of the rotating combination of double rotating plugs. Thus, in the in vessel fuel handling system of the double rotating plug type, it is necessary to decide the rotating plug size and evaluate the accessibility of in-vessel fuel handling machines in given core configuration. In this study, the simulation program based on LABVIEW which can effectively perform the arrangement design of the in vessel fuel handling system and simulate the rotating plug motion was developed. Fig. 1 shows the flow chart of the simulation program

  14. Pilot and pilot-commercial plants for reprocessing spent fuels of FBR type reactors

    International Nuclear Information System (INIS)

    Shaldaev, V.S.; Sokolova, I.D.

    1988-01-01

    A review of modern state of investigations on the FBR mixed oxide uranium-plutonium fuel reprocessing abroad is given. Great Britain and France occupy the leading place in this field, operating pilot plants of 5 tons a year capacity. Technology of spent fuel reprocessing and specific features of certain stages of the technological process are considered. Projects of pilot and pilot-commercial plants of Great Britain, France, Japan, USA are described. Economic problems of the FBR fuel reprocessing are touched upon

  15. Study of the Effect of Hydrocarbon Type Biodegradation on Fuel Specification Properties

    Science.gov (United States)

    2014-06-01

    diesel fuel (F10428) before and after 1 month exposure to Pseudomonas or a control. Figure 12. QCM profiles at 140°C of mass accumulation (solid...DLA-13) Figure 32. Calibration curve for analysis of BHT in jet fuel. Figure 33. Growth of yeast in 20 mg/L concentrations of A and B. Figure...bladder materials. Some costly problems associated with microbial growth include tank corrosion, fuel pump failures, filter plugging, injector

  16. Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2017-11-15

    In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

  17. Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

    International Nuclear Information System (INIS)

    Ferrufino, Felipe Bonito Jaldin

    2011-01-01

    IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)

  18. Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

    International Nuclear Information System (INIS)

    Permana, Sidik

    2009-01-01

    Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

  19. Fuel pin behavior under slow ramp-type transient-overpower conditions in the CABRI-FAST experiments

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, Jean

    2009-01-01

    In the CABRI-FAST experimental program, four in-pile tests were performed with slow power-ramp-type transient-overpower conditions (called hereafter as 'slow TOP') to study transient fuel pin behavior under inadvertent control rod withdrawal events in liquid metal cooled fast breeder reactors. Annular-pellet fuel pins were used in three tests, while a solid-pellet fuel pin was used in the other test. All of these pins were pre-irradiated in Phenix. The slow TOP test with a solid-pellet fuel pin was realized as a comparatory test against an existing test (E12) in the CABRI-2 program. In the CABRI-FAST test (BCF1), a power ramp rate of 3% Po/s was applied, while in the CABRI-2 test, 1% Po/s was adopted. Moreover, overpower condition was maintained for a few seconds beyond the observed pin failure in the BCF1 test. In spite of the different power ramp rates, evaluated fuel thermal conditions at the observed failure time are quite similar. The continued overpower condition in the BCF1 test resulted in gradual degradation of the pin structure providing information effective for evaluation of various accident scenarios. Three slow TOP tests with the annular fuel in the CABRI-FAST program resulted in no pin failure showing high failure threshold. Based on post-test examination data and a theoretical evaluation, it was concluded that intra-pin free spaces, such as central hole, macroscopic cracks and fuel-cladding gap effectively mitigated fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in case of very large amount of fuel melting. Furthermore, such cavity pressurization was effectively mitigated by a molten-fuel squirting into the upper blanket region pushing the blanket pellets upward. These CABRI FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database with various fuel and transient conditions. (author)

  20. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  1. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Uriarte, A; Ramos, L; Estrada, J; Val, J L. del

    1962-07-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO{sub 2}F{sub 2} solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs.

  2. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2003-01-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  3. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    Energy Technology Data Exchange (ETDEWEB)

    Mantecón, Javier González; Mattar Neto, Miguel, E-mail: javier.mantecon@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  4. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    International Nuclear Information System (INIS)

    Mantecón, Javier González; Mattar Neto, Miguel

    2017-01-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  5. Adsorptive on-board desulfurization over multiple cycles for fuel-cell-based auxiliary power units operated by different types of fuels

    Science.gov (United States)

    Neubauer, Raphael; Weinlaender, Christof; Kienzl, Norbert; Bitschnau, Brigitte; Schroettner, Hartmuth; Hochenauer, Christoph

    2018-05-01

    On-board desulfurization is essential to operate fuel-cell-based auxiliary power units (APU) with commercial fuels. In this work, both (i) on-board desulfurization and (ii) on-board regeneration performance of Ag-Al2O3 adsorbent is investigated in a comprehensive manner. The herein investigated regeneration strategy uses hot APU off-gas as the regeneration medium and requires no additional reagents, tanks, nor heat exchangers and thus has remarkable advantages in comparison to state-of-the-art regeneration strategies. The results for (i) show high desulfurization performance of Ag-Al2O3 under all relevant operating conditions and specify the influence of individual operation parameters and the combination of them, which have not yet been quantified. The system integrated regeneration strategy (ii) shows excellent regeneration performance recovering 100% of the initial adsorption capacity for all investigated types of fuels and sulfur heterocycles. Even the adsorption capacity of the most challenging dibenzothiophene in terms of regeneration is restored to 100% over 14 cycles of operation. Subsequent material analyses proved the thermal and chemical stability of all relevant adsorption sites under APU off-gas conditions. To the best of our knowledge, this is the first time 100% regeneration after adsorption of dibenzothiophene is reported over 14 cycles of operation for thermal regeneration in oxidizing atmospheres.

  6. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    International Nuclear Information System (INIS)

    Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-01-01

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  7. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-04-15

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  8. Comprehensive Analysis of Trends and Emerging Technologies in All Types of Fuel Cells Based on a Computational Method

    Directory of Open Access Journals (Sweden)

    Takaya Ogawa

    2018-02-01

    Full Text Available Fuel cells have been attracting significant attention recently as highly efficient and eco-friendly energy generators. Here, we have comprehensively reviewed all types of fuel cells using computational analysis based on a citation network that detects emerging technologies objectively and provides interdisciplinary data to compare trends. This comparison shows that the technologies of solid oxide fuel cells (SOFCs and electrolytes in polymer electrolyte fuel cells (PEFCs are at the mature stage, whereas those of biofuel cells (BFCs and catalysts in PEFCs are currently garnering attention. It does not mean, however, that the challenges of SOFCs and PEFC electrolytes have been overcome. SOFCs need to be operated at lower temperatures, approximately 500 °C. Electrolytes in PEFCs still suffer from a severe decrease in proton conductivity at low relative humidity and from their high cost. Catalysts in PEFCs are becoming attractive as means to reduce the platinum catalyst cost. The emerging technologies in PEFC catalysts are mainly heteroatom-doped graphene/carbon nanotubes for metal-free catalysts and supports for iron- or cobalt-based catalysts. BFCs have also received attention for wastewater treatment and as miniaturized energy sources. Of particular interest in BFCs are membrane reactors in microbial fuel cells and membrane-less enzymatic biofuel cells.

  9. Outward electron transfer by Saccharomyces cerevisiae monitored with a bi-cathodic microbial fuel cell-type activity sensor.

    Science.gov (United States)

    Ducommun, Raphaël; Favre, Marie-France; Carrard, Delphine; Fischer, Fabian

    2010-03-01

    A Janus head-like bi-cathodic microbial fuel cell was constructed to monitor the electron transfer from Saccharomyces cerevisiae to a woven carbon anode. The experiments were conducted during an ethanol cultivation of 170 g/l glucose in the presence and absence of yeast-peptone medium. First, using a basic fuel-cell type activity sensor, it was shown that yeast-peptone medium contains electroactive compounds. For this purpose, 1% solutions of soy peptone and yeast extract were subjected to oxidative conditions, using a microbial fuel cell set-up corresponding to a typical galvanic cell, consisting of culture medium in the anodic half-cell and 0.5 M K(3)Fe(CN)(6) in the cathodic half-cell. Second, using a bi-cathodic microbial fuel cell, it was shown that electrons were transferred from yeast cells to the carbon anode. The participation of electroactive compounds in the electron transport was separated as background current. This result was verified by applying medium-free conditions, where only glucose was fed, confirming that electrons are transferred from yeast cells to the woven carbon anode. Knowledge about the electron transfer through the cell membrane is of importance in amperometric online monitoring of yeast fermentations and for electricity production with microbial fuel cells. Copyright (c) 2009 John Wiley & Sons, Ltd.

  10. Gas chromatography-mass spectrometric determination of traces of ether-type icing inhibitors in free-floating fuels

    Energy Technology Data Exchange (ETDEWEB)

    Shin, H.S. [Dept. of Environmental Education, Kongju National Univ., Kongju (Korea); Abuse Drug Research Center, Kongju National Univ., Kongju (Korea); Ahn, H.S. [Dept. of Environmental Science, Kongju National Univ., Kongju (Korea)

    2004-08-01

    A gas chromatographic-mass spectrometric (GC-MS) assay method has been developed for simultaneous determination of ethylene glycol monomethyl ether (EGME) and diethylene glycol monomethly ether (DEGME) in spilled aviation fuels. Ethylene glycol monobutyl ether (EGBE) and ethylene glycol monoethyl ether (EGEE) were used as internal standard and surrogate, respectively. Sample preparation consisted of back-extraction with 7 mL dichloromethane after extraction of 50 mL of fuel with 2 mL of water. The extract was concentrated to dryness, dissolved in 100 {mu}L methanol, and analyzed by GC-MS with selected-ion monitoring (SIM). The peaks had good chromatographic properties on a semi-polar column. EGME and DEGME were extracted from fuel with high recovery of 75 and 85%, with small variations, respectively. Method detection limits were 1.3 and 1.0 ng mL{sup -1} for EGME and DEGME, respectively, in spilled fuel. DEGME was detected at concentrations of 22.6 and 19.7 ng mL{sup -1} in two samples from among five free-floating samples collected in a tunnel of a subway station located in the vicinity of an army base in Korea. The method might be useful for differentiation between the fuel-types kerosene and JP-8, which might originate from a storage tank. (orig.)

  11. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements

    International Nuclear Information System (INIS)

    Domingos, Douglas Borges

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a Miniplate Irradiation Device (MID) to be placed in the IEA-R1 reactor core. The irradiation device is used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 - Al dispersion fuels, LEU type (19.75 % 235 U) with uranium densities of, respectively, 3.2 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and 2DB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  12. Management and Handling of Rejected Fuel of MTR Type and Process Effluents Contained Uranium at FEPI

    International Nuclear Information System (INIS)

    Ghaib Widodo; Bambang Herutomo

    2007-01-01

    Research Reactor Fuel Element Production Installation (FEPI) - Serpong has performed management and handling of all kinds of rejected fuel material during production (solids, liquids, and gases) and process effluents contained uranium. The methods that has been implemented are precipitation, absorption, evaporation, electrolysis, and electrodialysis. By these methods will finally be obtained forms of product which can be used directly as fuel material feed and solid/liquid radioactive waste that fulfil the requirements (uranium contents < 50 ppm) to be send to Radioactive Waste Management Installation. (author)

  13. Core performance of equilibrium fast reactors for different coolant materials and fuel types

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Sekimoto, Hiroshi

    1998-01-01

    Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

  14. Investigation into the effect of different fuels on ignition delay of M-type diesel combustion process

    Directory of Open Access Journals (Sweden)

    Bibić Dževad

    2008-01-01

    Full Text Available An ignition delay is a very complex process which depends on a great number of parameters. In practice, definition of the ignition delay is based on the use of correlation expressions. However, the correlation expressions have very often limited application field. This paper presents a new correlation which has been developed during the research project on the direct injection M-type diesel engine using both the diesel and biodiesel fuel, as well as different values of a static injection timing. A dynamic start of injection, as well as the ignition delay, is defined in two ways. The first approach is based on measurement of a needle lift, while the second is based on measurement of a fuel pressure before the injector. The latter approach requires calculation of pressure signals delay through the fuel injection system and the variation of a static advance injection angle changing. The start of a combustion and the end of the ignition delay is defined on the basis of measurements of an in-cylinder pressure and its point of separation from a skip-fire pressure trace. The developed correlation gives better prediction of the ignition delay definition for the M-type direct injection diesel engine in the case of diesel and biodiesel fuel use when compared with the classic expression by the other authors available in the literature.

  15. Variation of power generation at different buffer types and conductivities in single chamber microbial fuel cells

    KAUST Repository

    Nam, Joo-Youn; Kim, Hyun-Woo; Lim, Kyeong-Ho; Shin, Hang-Sik; Logan, Bruce E.

    2010-01-01

    Microbial fuel cells (MFCs) are operated with solutions containing various chemical species required for the growth of electrochemically active microorganisms including nutrients and vitamins, substrates, and chemical buffers. Many different buffers

  16. Enhanced air/fuel mixing for automotive stirling engine turbulator-type combustors

    Science.gov (United States)

    Riecke, George T.; Stotts, Robert E.

    1992-01-01

    The invention relates to the improved combustion of fuel in a combustion chamber of a stirling engine and the like by dividing combustion into primary and secondary combustion zones through the use of a diverter plate.

  17. Radiation shielding at interim storage facility for CANDU-type nuclear spent fuel

    International Nuclear Information System (INIS)

    Mateescu, S.; Radu, M. Pantazi D.; Stanciu, M.

    1997-01-01

    Technical measures in radiological protection are taken in the interim storage facility design to ensure that, during normal operation, exposures of workers and members of public to ionizing radiation are limited to levels lower than regulatory limits. The spent fuel storage design provides for radiation exposure to be as low as reasonable achievable (ALARA principles). The evaluation of radiation shields includes the most conservative provisions: - all locations which may contain spent fuel are full; - the spent fuel has reached the maximum burnup; - the post irradiation cooling period should be the minimum reasonable; - equipment for handling contains the maximum amount of spent fuel. Radiation shields should ensure that external radiation fields do not exceed limits accepted by the Regulatory Body Module. The evaluation has been performed with two computer codes, QAD-5K and MICROSHIELD-4. (authors)

  18. Study of core characteristics on fuel and coolant type. Results of F/S phase-I

    International Nuclear Information System (INIS)

    Ikegami, Tetsuo; Hayashi, Hideyuki; Sasaki, Makoto; Mizuno, Tomoyasu; Yamadate, Megumi; Takaki, Naoyuki; Kurosawa, Norifumi; Sakashita, Yoshiaki; Naganuma, Masayuki

    2001-03-01

    The phase-I of the Feasibility Study of Commercialized Fast Reactor Cycle Systems (F/S) were started from July, 1999 and terminated at the end of FY2000 in order to executed examination about technology alternatives of various commercialized fast reactor (FR) recycle concepts, in response to the JNC middle long term enterprise plan. In the phase-I of this F/S, a number of conceptual candidates have been selected from the following 5 viewpoints: a) ensuring safety, b) economic competitiveness to future LWRs, c) efficient utilization of resources, d) reduction of environmental burden, e) enhancement of nuclear non-proliferation. As for this study from the above viewpoints, core characteristics of many kinds of reactors have been investigated, analyzed and examined a core / a fuel characteristic in the combinations of fuel and coolant types and power output scales. Based on these results, R and D plans of the phase-II to be performed have been proposed, and a database to select candidate reactor concepts has been prepared. The conclusions have been obtained in the phase-I are as follows: (1) Evaluation of a fuel form in every each coolant was compared. A promising fuel form was extracted as follows: an oxide and a metal fuel for sodium coolant cores, a metal and a nitride fuel for heavy metal coolant cores, an oxide and a nitride fuel for carbon dioxide coolant cores and a nitride fuel for He gas coolant cores. (2) As the general idea that performance of a core nucleus can be compatible with re-criticality evasion in sodium coolant large-sized oxide fuel cores, a axial blanket particle elimination radial heterogeneous core is one influential candidate. (3) In case of Pb-Bi coolant nature circulation medium size core with an oxide fuel, it is difficult to simultaneously achieve higher discharged burn-up and higher breeding ratio according to the viewpoints of the phase-I. (4) Core characteristics of a carbon dioxide coolant core shows to be almost equivalent to that of

  19. DANCOFF-MC: A program to calculate Dancoff factors in CANDU type fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S; Valko, J

    1992-12-01

    The objective of DANCOFF-MC is the evaluation of Dancoff factors for cylindrical fuel rods arranged parallel in various and complicated bundle geometries. No interaction with fuel rods in any of the other bundles are considered due to the large distance, in mean free paths, between the buldes. Using a common basic algorithm three versions of the program have been written so far: The DANCOFF-MC-2, the DANCOFF-MC-19 and the DANCOFF-MC-27. (orig./HP).

  20. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Daverio, Hernando J.

    2003-01-01

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  1. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    1988-06-01

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

  2. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  3. Characterization of dispersed type fuel miniplates based in alloy UMo by evaluation of changes volumetrics and thermal conductivity

    International Nuclear Information System (INIS)

    Salinas Valero, Pablo Ignacio

    2016-01-01

    The development of new technologies in the nuclear area is extremely important to achieve greater efficiency and security in the production of electrical energy in the case of power reactors and for the production of radioisotopes and neutrons in research reactors. Throughout history, uranium-based nuclear fuels evolved in parallel with the requirements of nuclear reactors, this emphasis was increased when the RERTR program was created, which restricts the use of fuels with a maximum enrichment of 20% of the isotope U 235 (fissile isotope), which makes it necessary to increase the mass of uranium to compensate the amount of fissile material to maintain a neutron flux necessary for the reactors to operate with the same power. The search for new nuclear fuels has reached the UMo alloy with which densities of 18 gU/cm 3 are achieved in type fuels and 8 gU/cm 3 in dispersed type fuels, properties under irradiation due to their cubic crystalline structure. This type of fuel, when used dispersed in an aluminum matrix, becomes thermodynamically unstable by increasing the fission temperature of the U 235 isotope, due to this, compounds of lower density are formed, which causes an increase in volume (swelling). ). This swelling is studied throughout the present work, to relate the changes of UMo-Al / 4% volume of thermally induced miniecography in thermal treatments, with the purpose of evaluating changes in the thermal conductivity of the material. In this study it was detected that the swelling in miniplates is related in some way to the reduction of thermal conductivity, it was also recorded that the volume of change is irregular increasing and decreasing its volume according to the hours of induced swelling. The purpose of this work is to contribute to the development of dispersed fuels based on the UMo alloy in order to control the variables and reduce the probability of faults and possible accidents, such as fission products, or an increase in temperature in the core

  4. The use of U3Si2 dispersed in aluminum in plate-type fuel elements for research and test reactors

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U 3 Si 2 dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U 3 Si 2 fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U 3 Si 2 particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U 3 Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U 3 Si 2 -aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m 3 is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs

  5. Reactivity worth of the thermal column of a MTR type swimming pool research reactor using low enriched uranium fuel

    International Nuclear Information System (INIS)

    Ali Khan, L.; Ahmad, N.

    2002-01-01

    The reactivity worth of the thermal column of a typical MTR type swimming pool research reactor using low enriched uranium fuel has been determined by modeling the core using standard computer codes. It was also measured experimentally by operating the reactor in the stall and open ends. The calculated value of the reactivity worth of the thermal column is about 14% greater than the experimentally determined value

  6. Report on swelling of MX-type fuels 1973/76: Self-diffusion in MX-type nuclear fuels out-of-pile and in-pile

    International Nuclear Information System (INIS)

    Matzke, H.; Bradbury, M.H.

    1978-01-01

    Self-diffusion measurements of Pu-238 and U-233 have been carried out in a wide range of advanced nuclear fuels in the temperature region from 1200 to 2300 0 C. The materials studied varied in composition from carbides through carbonitrides to nitrides. In particular the effect on self-diffusion rates of factors such as non-metal/metal ratio, oxygen content increasing nitrogen contents, metallic impurity additions, the presence of second phases and fission products simulating 16, 10 and 3 a /o burn up has been established. Grain boundary diffusion rates were evaluated where possible. Carbon diffusion in stoichiometric and off-stoichiometric UC and in a series of uranium carbonitride samples was also measured. The RADIF experiments (radiation induced diffusion) have provided results upon the effect of irradiation on the self-diffusion rates in the temperature range 150 to 1300 0 C. Each of the factors mentioned above is discussed in detail with special attention being given to the effects of non-metal/metal ratio, impurities and increasing the nitrogen content in carbonitride materials

  7. GUIA METODOLOGICA PARA UN MANEJO INTEGRAL COSTERO APLICADO A PEHUEN CO (ARGENTINA

    Directory of Open Access Journals (Sweden)

    María Luján Bustos

    2017-03-01

    Os planos de gestão costeira integrada (GCI na Argentina são quase inexistentes. No Sul da província de Buenos Aires (Argentina, o planejamento por causa da intensidade da erosão costeira, que afeta especialmente as áreas urbanizadas, é necessário. É fundamental ainda traçar estratégias de GCI que possam ser aplicadas em toda a área de forma sustentável. Neste processo, a participação e o compromisso dos atores sociais são fundamentais. Por esta razão, um guia metodológico foi construído para uma GCI com participação social e aplica-se à localidade de Pehuen Co. Obteve-se como resultado que a erosão costeira e a sazonalidade do turismo são os pilares tanto para a visão científica quanto para a social. Sobre isso, foram delineadas ações e estratégias a curto, médio e longo prazos e indicadores de acompanhamento. Concluiu-se que é necessário o compromisso das autoridades locais, trabalhando em conjunto com as partes interessadas, para o bom funcionamento do GCI. Palavras-chave: Gestão Integrada Costeira; Indicadores Ambientais; Erosão; Sazonalidade do Turismo; Pehuen Co (Argentina.

  8. Exame físico na criança: um guia para o enfermeiro

    Directory of Open Access Journals (Sweden)

    Josilene de Melo Buriti Vasconcelos

    1999-12-01

    Full Text Available O exame físico constitui-se numa importante etapa dentro do processo de avaliação à criança, através do qual o enfermeiro enriquece as informações obtidas durante a entrevista e fundamenta a assistência de enfermagem. Na atualidade, mediante o crescente interesse dos enfermeiros em aplicar o processo de enfermagem em todas as suas etapas, o exame físico tem ocupado lugar de destaque, por permitir que os enfermeiros conheçam as necessidades de seus clientes, no que diz respeito aos seus aspectos físicos e fisiológicos, permitindo a identificação dos diagnósticos de enfermagem, alem de servir como recurso para avaliação efetiva das intervenções de enfermagem. Este estudo, de caráter bibliográfico, objetivou elaborar um guia para realização do exame físico na criança, a ser utilizado por enfermeiros, durante o processo de avaliação à criança. Nele apresentamos o exame físico de forma sistemática, utilizando a seqüência céfalo-caudal, descrevendo os possíveis achados normais e anormais ao examinarmos a criança.

  9. Ver, Compreender e Aprender: estudo comparado com guias de uso de apps editoriais para dispositivos móveis

    Directory of Open Access Journals (Sweden)

    Ana Claudia Dalagnoli

    2016-12-01

    Full Text Available Este estudo foi construído a partir da percepção de que a evolução das tecnologias e, sobretudo, o aprimoramento da comunicação em dispositivos móveis, impacta diretamente no processo de desenvolvimento de aplicativos (apps e softwares e na relação do ser humano com as novas interfaces. A pesquisa apresenta uma análise comparativa entre os aplicativos para iPad das revistas de decoração e design de interiores Casa Claudia (Editora Abril e Casa e Jardim (Editora Globo, a partir de testes de usabilidade dos mesmos em tablets Apple. O objetivo foi verificar se os guias de uso disponíveis em aplicativos (apps de revistas para dispositivos móveis, construídos por textos e imagens, estão cumprindo a função de comunicar, efetivamente, o modo de uso desses apps. As informações foram coletadas junto a mulheres com idade entre 25 e 54 anos que testaram e comentaram o funcionamento das versões atuais dos apps. Foram selecionadas 24 mulheres para fazerem atividades individuais monitoradas em ambos os apps, com dados coletados por meio de observações e think aloud. Os dados obtidos foram analisados, e com base nos resultados foi proposto um ajuste nos requisitos para o desenvolvimento de interfaces de guias para aplicativos com vistas à maior satisfação das usuárias. Foram percebidas falhas severas na apreensão e aplicação das orientações disponíveis nos guias de uso. Os pesquisadores verificaram que os requisitos básicos para aumentar a satisfação e melhorar a usabilidade estão relacionados à facilidade de compreensão e à simplicidade no uso, o que pode ser obtido por meio de melhorias relacionadas à aplicação da tecnologia, ao conhecimento dos processos cognitivos dos usuários e à aplicação de conceitos semióticos de comunicação. O estudo pode servir de base para o desenvolvimento e análise de guias de uso de outros apps com características similares, em diversos sistemas operacionais.

  10. Adaptação multicêntrica do guia para a gestão autônoma da medicação

    Directory of Open Access Journals (Sweden)

    Rosana Teresa Onocko Campos

    2012-12-01

    Full Text Available O uso crescente de psicofármacos e o baixo empowerment dos usuários mostram-se críticos à qualificação da assistência em Saúde Mental no Brasil. Este estudo, abrangendo três cidades brasileiras, objetivou a elaboração do Guia Brasileiro da Gestão Autônoma da Medicação (GGAM-BR, com base na tradução e adaptação de guia desenvolvido no Canadá; e a avaliação dos efeitos do uso do GGAM-BR na formação de trabalhadores de saúde mental. Constituíram-se grupos de intervenção (GIs para compartilhamento das experiências com tratamento medicamentoso, a partir dos temas propostos no guia; e foram realizados grupos focais antes e após os GIs. Importantes mudanças em relação ao texto original do guia Canadense foram implementadas, levando em conta a realidade brasileira. Constatou-se que o GGAM-BR constitui estratégia potente de fomento à participação ativa dos usuários na gestão do tratamento e do serviço, incidindo positivamente na formação de trabalhadores.

  11. Trapped in the heat: A post-communist type of fuel poverty

    International Nuclear Information System (INIS)

    Tirado Herrero, Sergio; Ürge-Vorsatz, Diana

    2012-01-01

    Fuel poverty is a still insufficiently researched social and energy challenge with significant climate change implications. Based on evidence from Hungarian panel apartment blocks connected to district heating, this paper introduces a new variant of fuel poverty that may not be properly captured by existing fuel poverty indicators. This newly defined variant can be largely attributed to post-communist legacies – though it might also exist in other contexts – and assumes that consumers living in poor-efficiency, district-heated buildings are trapped in dwellings with adequate indoor temperatures but disproportionately high heating costs because (a) changing supplier or fuel is difficult because of the existing technical and institutional constraints, and (b) they do not realistically have the option to reduce individually their heating costs through individual efficiency improvements. This situation often translates into payment arrears, indebtedness, risk of disconnection, or reduced consumption of other basic goods and services. State-supported policy responses to date have favoured symptomatic solutions (direct consumer support) combined with superficial retrofits, though it is argued that only state-of-the-art retrofits such as the passive house-based SOLANOVA pilot project in Dunaújváros can fully eradicate fuel poverty in this consumer group. - Highlights: ► We identify a new variant of fuel poverty. ► We explore this variant in panel apartment blocks connected to DH in Hungary, where dwellings are warm enough in winter but have disproportionately high energy costs. ► Affected households react in ways that harm their welfare and put them at risk. ► Deep retrofits in dwellings such as these can eradicate fuel poverty while also contributing to other goals.

  12. Online failed fuel identification using delayed neutron detector signals in pool type reactors

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

    2011-01-01

    In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

  13. Laboratory measurements of trace gas emissions from biomass burning of fuel types from the southeastern and southwestern United States

    Science.gov (United States)

    Burling, I. R.; Yokelson, R. J.; Griffith, D. W. T.; Johnson, T. J.; Veres, P.; Roberts, J. M.; Warneke, C.; Urbanski, S. P.; Reardon, J.; Weise, D. R.; Hao, W. M.; de Gouw, J.

    2010-11-01

    Vegetation commonly managed by prescribed burning was collected from five southeastern and southwestern US military bases and burned under controlled conditions at the US Forest Service Fire Sciences Laboratory in Missoula, Montana. The smoke emissions were measured with a large suite of state-of-the-art instrumentation including an open-path Fourier transform infrared (OP-FTIR) spectrometer for measurement of gas-phase species. The OP-FTIR detected and quantified 19 gas-phase species in these fires: CO2, CO, CH4, C2H2, C2H4, C3H6, HCHO, HCOOH, CH3OH, CH3COOH, furan, H2O, NO, NO2, HONO, NH3, HCN, HCl, and SO2. Emission factors for these species are presented for each vegetation type burned. Gas-phase nitrous acid (HONO), an important OH precursor, was detected in the smoke from all fires. The HONO emission factors ranged from 0.15 to 0.60 g kg-1 and were higher for the southeastern fuels. The fire-integrated molar emission ratios of HONO (relative to NOx) ranged from approximately 0.03 to 0.20, with higher values also observed for the southeastern fuels. The majority of non-methane organic compound (NMOC) emissions detected by OP-FTIR were oxygenated volatile organic compounds (OVOCs) with the total identified OVOC emissions constituting 61 ± 12% of the total measured NMOC on a molar basis. These OVOC may undergo photolysis or further oxidation contributing to ozone formation. Elevated amounts of gas-phase HCl and SO2 were also detected during flaming combustion, with the amounts varying greatly depending on location and vegetation type. The fuels with the highest HCl emission factors were all located in the coastal regions, although HCl was also observed from fuels farther inland. Emission factors for HCl were generally higher for the southwestern fuels, particularly those found in the chaparral biome in the coastal regions of California.

  14. Improvement of visualization efficiency for the nondestructive inspection image of internal defects in plate type nuclear fuel

    International Nuclear Information System (INIS)

    Park, Seung Kyu; Park, Nak Kyu; Baik, Sung Hoon; Lee, Yoon Sang; Cheong, Yong Moo; Kang, Young June

    2012-01-01

    Plate type nuclear fuel has been adopted in most research reactors. The production quality of the fuel is a key part for an efficient and stable generation of thermal energy in research reactors. Thus, a nondestructive quality inspection for the internal defects of plate type nuclear fuel is a key process during the production of nuclear fuel for safety insurance. Nondestructive quality inspections based on X rays and ultrasounds have been widely used for the defect detection of plate type nuclear fuel. X ray testing is a simple and fast inspection method, and provides an image in real time as the inspection results. Thus, the testing can be carried out by a non expert field worker. However, it is hard to detect closed type defects that should be detected during the production of plate type nuclear fuel. Ultrasonic testing is a powerful tool to detect internal defects including open type and closed type defects in plate type nuclear fuel. However, the inspection process is complicated because an immersion test should be carried out in a water tank. It is also a time consuming inspection method because area testing to acquire image is based on the scanning of the point by point inspections. Among nondestructive inspection techniques, the techniques based on laser interferometry and infrared thermography have been widely used in the detection of internal defects of plate type composite materials, such as aircraft, automotive etc. While infrared thermography technique (IRT) analyses the thermal behavior of the specimen surface, laser interferometry technique (LIT) analyses the deformation field. Both techniques are useful tools for detection and evaluation of internal defects in composite materials. Especially, the laser interferometry technique can provide the depth information of internal defects. Laser interferometry technique (LIT) is a non contact inspection method faster than thermography. Also, this technique requires less energy than thermography and the

  15. Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

    2009-08-01

    Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

  16. Economic evaluation of dual purpose desalination plants by fuel type in Korea

    International Nuclear Information System (INIS)

    Seung-Su, Kim; Man-Ki, Lee

    2007-01-01

    In light of the recent rapid increase in the fossil fuel prices it is meaningful to evaluate the impact of these price changes in the economics of dual-purpose desalination projects producing electricity and fresh water simultaneously. The price of crude oil and LNG (Liquefied Natural Gas) has increased by about 200% and 100% during the past three or four years. The uranium price has also increased by nearly 500% during the same period. The purpose of this paper is to analyze and compare the economics of SMART (System-integrated Modular Advanced ReacTor) which is being developed as a small size PWR type and the LNG Combine Cycle coupled with MED (Multi-Effect Distillation) which are being acknowledged as promising energy sources for the future in Korea. The methods of analysis used in this paper are the lifetime leveled cost method for the power and water cost calculation and the power credit method for the total cost allocation. DEEP (Devaluation Economic Evaluation Program) developed by IAEA was used to perform an economic comparison between the two dual-purpose desalination projects. From the results of the analysis it is found that the desalination by SMART-MED is much superior to that of LNG CC-MED under the current economic and technical situations. It is also shown that the relative superiority of SMART-MED to LNG CC-MED will be maintained even in case where an increase in the uranium price and the SMART construction cost would reach a maximum in the sensitivity analysis. In the case that the discount rate declines to 5% per year, the relative attractiveness of SMART-MED which is a capital intensive plant will be enhanced when compared to that for a 7% discount rate. In addition to this, it is thought that a nuclear energy source will be favored much more than now in the field of desalination if the regulations for the emission of greenhouse gases are to be strengthened. (authors)

  17. Ab initio study of perovskite type oxide materials for solid oxide fuel cells

    Science.gov (United States)

    Lee, Yueh-Lin

    2011-12-01

    Perovskite type oxides form a family of materials of significant interest for cathodes and electrolytes of solid oxide fuel cells (SOFCs). These perovskites not only are active catalysts for surface oxygen reduction (OR) reactions but also allow incorporating the spilt oxygen monomers into their bulk, an unusual and poorly understood catalytic mechanism that couples surface and bulk properties. The OR mechanisms can be influenced strongly by defects in perovskite oxides, composition, and surface defect structures. This thesis work initiates a first step in developing a general strategy based on first-principles calculations for detailed control of oxygen vacancy content, transport rates of surface and bulk oxygen species, and surface/interfacial reaction kinetics. Ab initio density functional theory methods are used to model properties relevant for the OR reactions on SOFC cathodes. Three main research thrusts, which focus on bulk defect chemistry, surface defect structures and surface energetics, and surface catalytic properties, are carried to investigate different level of material chemistry for improved understanding of key physics/factors that govern SOFC cathode OR activity. In the study of bulk defect chemistry, an ab initio based defect model is developed for modeling defect chemistry of LaMnO 3 under SOFC conditions. The model suggests an important role for defect interactions, which are typically excluded in previous defect models. In the study of surface defect structures and surface energetics, it is shown that defect energies change dramatically (1˜2 eV lower) from bulk values near surfaces. Based on the existing bulk defect model with the calculated ab initio surface defect energetics, we predict the (001) MnO 2 surface oxygen vacancy concentration of (La0.9Sr0.1 )MnO3 is about 5˜6 order magnitude higher than that of the bulk under typical SOFC conditions. Finally, for surface catalytic properties, we show that area specific resistance, oxygen

  18. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    International Nuclear Information System (INIS)

    Solyany, V.I.; Bibilashvili, Yu.K.; Sukhanov, G.I.; Pimenov, Yu.V.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-01-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness. (author)

  19. PARAMETERS OF AIR FIRED BOILER FED WITH DIFFERENT TYPES OF FUEL

    Directory of Open Access Journals (Sweden)

    Katarzyna Joanna Gładyszewska-Fiedoruk

    2016-09-01

    Full Text Available The measurement and interpretation of indoor carbon dioxide CO2 concentration can provide information on building indoor air quality and ventilation. On the other hand, concentration of carbon monoxide CO can show as how combustion process run and if the boiler is safe. When there is not sufficient air available to complete the combustion process, some of the fuel is left unburned, resulting in inefficiency and undesirable emissions. An examination of the CO2 and CO concentration in boiler and interpretation results help to improve indoor air quality. The paper presents characteristics of concentration CO2 and CO depend on used fuel in tested boiler rooms. The concentration curves show how each fuel combustion affect the amount of CO2 and CO that is produced.

  20. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Solyany, V I; Bibilashvili, Yu K; Sukhanov, G I; Pimenov, Yu V [Vsesoyuznyj Nauchno-Issledovatel' skij Inst. Neorganicheskikh Materialov, Moscow (USSR); Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-12-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness.

  1. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR

    International Nuclear Information System (INIS)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L.

    2008-01-01

    In this work was developed the neutronic model of a fuel sphere of a nuclear reactor of gas of high temperature to modulate of bed of spheres (PBMR), using the Monte Carlo method with the MCNPx code. In order to be able to verify the fuel model constructed in this investigation, it is used a case of reference, based on an international exercise b enchmark . The benchmark report contains the results sent by different international participants for five phases with respect to the high temperature gas reactor (HTR), fed with uranium, plutonium and thorium. In particular, in first stage of benchmark an infinite adjustment of uranium compound fuel spheres is considered unique, with which our results were compared. This first stage considers two cases: cell calculations with spherical external frontier and cell calculations with cubic external frontier. The objective is to identify any increase in the uncertainty, related to the uranium fuel, that is associated with the plutonium and thorium fuels. In order to validate our results, the values of the neutron multiplication factor were taken in account, in cold and in the heat of the moment from the participants who sent their results obtained with Monte Carlo and deterministic calculations. The model of the fuel sphere developed in this work considers a regular distribution of 15000 Triso particles, in a cubic mesh centered within the sphere. For it was necessary to define the step firstly or p itch o f the cubic mesh. Generally, the results obtained by the participants of benchmark and those of this investigation present good agreement, nevertheless, appear some discrepancies, attributed to factors like different libraries of cross sections used, the nature of the solution: Monte Carlo or deterministic, and the difficulty of some participants to model the external frontier condition of reflection. (Author)

  2. The encapsulation of Magnox type fuel elements for extended storage in cooling ponds

    International Nuclear Information System (INIS)

    Baker, D.W.C.; Burt, G.A.

    1978-01-01

    A method of encapsulating spent fuel elements in a protective plastics medium to enable them to be stored for protracted periods under water, without risk of further significant corrosion, has been developed. It is visualised that the elements after discharge from the reactor would be allowed to cool under water for a period of at least 100 days and would then be encapsulated while remaining immersed. A suitable two pack system based on a solvent free epoxy resin cured with an aromatic amine adduct has been identified. The equipment and processes which have been developed for handling, conditioning and encapsulating the fuel are described. (author)

  3. A calculation methodology applied for fuel management in PWR type reactors using first order perturbation theory

    International Nuclear Information System (INIS)

    Rossini, M.R.

    1992-01-01

    An attempt has been made to obtain a strategy coherent with the available instruments and that could be implemented with future developments. A calculation methodology was developed for fuel reload in PWR reactors, which evolves cell calculation with the HAMMER-TECHNION code and neutronics calculation with the CITATION code.The management strategy adopted consists of fuel element position changing at the beginning of each reactor cycle in order to decrease the radial peak factor. The bi-dimensional, two group First Order perturbation theory was used for the mathematical modeling. (L.C.J.A.)

  4. Fuel consumption: short term and long term price impacts per population type

    International Nuclear Information System (INIS)

    2011-01-01

    This report presents assessments of the price sensitivity of household fuel consumption. After a literature review on price-elasticity assessments and the use of pseudo-panels, the investigation analyses the deciding factors of the household fuel expense and its evolution between 1985 and 2006. It proposes a short term price-elasticity assessment based on the most recent survey, and also proposes price-elasticity assessments for sub-populations, notably in terms of income level or location (rural or urban areas)

  5. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    OpenAIRE

    Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

    2016-01-01

    Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

  6. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  7. Computational comparison of the effect of mixing grids of 'swirler' and 'run-through' types on flow parameters and the behavior of steam phase in WWER fuel assemblies

    International Nuclear Information System (INIS)

    Shcherbakov, S.; Sergeev, V.

    2011-01-01

    The results obtained using the TURBOFLOW computer code are presented for the numerical calculations of space distributions of coolant flow, heating and boiling characteristics in WWER fuel assemblies with regard to the effect of mixing grids of 'Swirler' and 'Run-through' types installed in FA on the above processes. The nature of the effect of these grids on coolant flow was demonstrated to be different. Thus, the relaxation length of cross flows after passing a 'Run-through' grid is five times as compared to a 'Swirler'-type grid, which correlates well with the experimental data. At the same time, accelerations occurring in the flow downstream of a 'Swirler'-type grid are by an order of magnitude greater than those after a 'Run-through' grid. As a result, the efficiency of one-phase coolant mixing is much higher for the grids of 'Run-through' type, while the efficiency of steam removal from fuel surface is much higher for 'Swirler'-type grids. To achieve optimal removal of steam from fuel surface it has been proposed to install into fuel assemblies two 'Swirler'-type grids in tandem at a distance of about 10 cm from each other with flow swirling in opposite directions. 'Run-through' grids would be appropriate for use for mixing in fuel assemblies with a high non-uniformity of fuel-by-fuel power generation. (authors)

  8. Study on the Applicability of Electron Beam Welding Methods to Assembly a Fuel Compact and Al Cover Plate of Research Reactor Plate Type Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae In; Lee, Yoon Sang; Lee, Don Dae; Jeong, Yong Jin; Kwon, Sun Chil; Kim, Soo Sung; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Among the research reactor plate type fuel fabrication processes, there is an assembly process between fuel meat compact and Al cover plates using a welding method prior to rolling process. The assembly process is such as the Al frame and Al cover plate should be welded properly as shown in Fig. 1. For welding, TIG(Tungsten Inert Gas) welding methods has been used conventionally, but in this study an electron beam welding(EB welding) technique which uses the electron beam of a high velocity for joining two materials is introduced to the assembly. The work pieces are melted as the kinetic energy of the electron beam is transformed into heat to join the two parts of the weld. The welding is often done in the conditions in a vacuum to prevent dispersion of the electron beam. The electron beam welding process has many ad-vantages such as contamination of the welds could be prevented, the penetration of the weld is deep, and also the strain of the welding area is less than other methods. In this study, to find optimal condition of the EB welding process, a welding speed, a beam current and an acceleration voltage were changed. To analyzing the welding results, the shape of the beads and defects of welding area was used. The width and depth of the beads were measured as well

  9. Study on the Applicability of Electron Beam Welding Methods to Assembly a Fuel Compact and Al Cover Plate of Research Reactor Plate Type Fuel

    International Nuclear Information System (INIS)

    Lee, Hae In; Lee, Yoon Sang; Lee, Don Dae; Jeong, Yong Jin; Kwon, Sun Chil; Kim, Soo Sung; Park, Jong Man

    2012-01-01

    Among the research reactor plate type fuel fabrication processes, there is an assembly process between fuel meat compact and Al cover plates using a welding method prior to rolling process. The assembly process is such as the Al frame and Al cover plate should be welded properly as shown in Fig. 1. For welding, TIG(Tungsten Inert Gas) welding methods has been used conventionally, but in this study an electron beam welding(EB welding) technique which uses the electron beam of a high velocity for joining two materials is introduced to the assembly. The work pieces are melted as the kinetic energy of the electron beam is transformed into heat to join the two parts of the weld. The welding is often done in the conditions in a vacuum to prevent dispersion of the electron beam. The electron beam welding process has many ad-vantages such as contamination of the welds could be prevented, the penetration of the weld is deep, and also the strain of the welding area is less than other methods. In this study, to find optimal condition of the EB welding process, a welding speed, a beam current and an acceleration voltage were changed. To analyzing the welding results, the shape of the beads and defects of welding area was used. The width and depth of the beads were measured as well

  10. Influence of hydrogen contamination by mercury on the lifetime of the PEM-type fuel cell

    Czech Academy of Sciences Publication Activity Database

    Bouzek, K.; Paidar, M.; Mališ, J.; Jakubec, Ivo; Janík, L.

    2010-01-01

    Roč. 56, č. 2 (2010), s. 889-895 ISSN 0013-4686 Institutional research plan: CEZ:AV0Z40320502 Keywords : fuel cell * power output * hydrogen contamination Subject RIV: CG - Electrochemistry Impact factor: 3.642, year: 2010

  11. Fundamental principles of failed fuel detection concepts on nuclear power units of WWER type

    International Nuclear Information System (INIS)

    Lusanova, L.; Miglo, V.; Slavyagin, P.

    2001-01-01

    The subject of the paper is the Russian failed fuel detection concept in both operating and shut down reactors. The philosophy for detection of fission products released from defective fuel during operation and sipping tests and using of these results for regulation of the radiological situation at the NPP during the next cycle is widely spread. In presented work such philosophy is applied to the shut down rectors. An option for sipping test performed in a mast of Refueling Machine (RM) using a wet-gas version of sipping test is briefly described. The use of the FFD method in RM mast allows combining the procedure of Fuel Assemblies (FA) tightness test with transport operation during reloading of the fuel from the core into the cooling pool. This reduces the time for reloading and transport operation with FA and increases the safety of reactor operation. The FFD method in RM mast has passed successful tests on Unit 4 at Balakovskaja NPP and it is expected to apply in other NPP unit with WWER-1000 reactors

  12. 40 CFR 600.207-86 - Calculation of fuel economy values for a model type.

    Science.gov (United States)

    2010-07-01

    ... calculation of the original base level fuel economy values), and (iii) All subconfigurations within the new... a new base level. The new base level is identical to the existing base level except that it shall be considered, for the purposes of this paragraph, as containing a new basic engine. The manufacturer will be...

  13. 40 CFR 600.207-93 - Calculation of fuel economy values for a model type.

    Science.gov (United States)

    2010-07-01

    ... calculation of the original base level fuel economy values); and (iii) All subconfigurations within the new... a new base level. The new base level is identical to the existing base level except that it shall be considered, for the purposes of this paragraph, as containing a new basic engine. The manufacturer will be...

  14. 14 CFR 26.35 - Changes to type certificates affecting fuel tank flammability.

    Science.gov (United States)

    2010-01-01

    ... assessment of the fuel tank system, as modified by their design change. The assessment must identify any... and applicants subject to paragraph (a)(1) or (a)(3)(iii) of this section, if the assessment required... tanks. (c) Impact Assessment. By the times specified in paragraphs (c)(1) and (c)(2) of this section...

  15. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Wang Qiming; Yan Xiaoqing [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Ding Shurong, E-mail: dsr1971@163.co [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

    2010-04-01

    The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

  16. Transport dynamics of a high-power-density matrix-type hydrogen-oxygen fuel cell

    Science.gov (United States)

    Prokopius, P. R.; Hagedorn, N. H.

    1974-01-01

    Experimental transport dynamics tests were made on a space power fuel cell of current design. Various operating transients were introduced and transport-related response data were recorded with fluidic humidity sensing instruments. Also, sampled data techniques were developed for measuring the cathode-side electrolyte concentration during transient operation.

  17. The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

    International Nuclear Information System (INIS)

    Xiong, Zhenqin; Gu, Hanyang; Wang, Minglu; Cheng, Ye

    2014-01-01

    Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m 2 /s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10 −2 m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m 2 /s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow rate and

  18. The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Xiong, Zhenqin [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Gu, Hanyang, E-mail: guhanyang@stu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Wang, Minglu [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Cheng, Ye [Shanghai Nuclear Engineering Research and Design Institute, Shanghai 200233 (China)

    2014-12-15

    Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m{sup 2}/s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10{sup −2} m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m{sup 2}/s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow

  19. Study on a New Type of Electric-controlled Engine Fuel Consumption Meter Based on Volume Method

    Directory of Open Access Journals (Sweden)

    Qing-Yong Zhang

    2014-04-01

    Full Text Available At present study on the testing methods and instruments for vehicles’ fuel consumption is still not perfect. It still can’t provide a rapid and accurate measuring method and instrument. A new type of fuel consumption meter structure is developed which used two small containers to relay to supply the engine and realizes oil consumption measuring by detecting the real- time liquid level in the containers. Photoelectric sensors and a chip microcomputer are used to realize transient detection. Its structure and principle are analyzed. The system of its hardware and software of the electric-controlling system are designed. Some key components are selected and the process of exhausting, starting and measuring are designed. Precision test of the system is performed, and the result shows the accuracy of the meter in the range of 800 ml is 0.1 %, which meets the requirements and the feasibility of the structure is verified. Finally the main influencing factors are analyzed.

  20. Experimental evaluation of the wear of the PEC type fuel element base. Tribological experimental studies in Na at high temperature

    International Nuclear Information System (INIS)

    D'Agraives, B.C.; Volcan, A.; Bacchilega, A.

    1978-01-01

    Tribological studies in sodium, related to the PEC-type fuel element design are presented. They are aimed at the simulation of friction and wear phenomena which are expected to occur on the surface of fuel element components undergoing solid-solid contact situations with variable loads and/or variable motions. In this first paper, a description of the preparatory work is given. Then, results related to long-duration experiments are shown with respect to the contact between the centering spherical ring belonging to the lower extension of the subassembly, and the cylindrical sleeve of the grid in which it takes place. After 1000 hours under loaded and vibrated conditions, in sodium at 400 0 C, the wear effects suffered by both contacting samples, are observed and evaluated. The stellite surfaces of the samples are damaged to a not-negligible extent and material transfers from the cylindrical sleeve onto the spherical ring occur

  1. Detection of delamination defects in plate type fuel elements applying an automated C-Scan ultrasonic system

    International Nuclear Information System (INIS)

    Katchadjian, P.; Desimone, C.; Ziobrowski, C.; Garcia, A.

    2002-01-01

    For the inspection of plate type fuel elements to be used in Research Nuclear Reactors it was applied an immersion pulse-echo ultrasonic technique. For that reason an automated movement system was implemented according to the axes X, Y and Z that allows to automate the test and to show the results obtained in format of C-Scan, facilitating the immediate identification of possible defects and making repetitive the inspection. In this work problems found during the laboratory tests and factors that difficult the inspection are commented. Also the results of C-Scans over UMo fuel elements with pattern defects are shown. Finally, the main characteristics of the transducer with the one the better results were obtained are detailed. (author)

  2. The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

    International Nuclear Information System (INIS)

    Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

    1995-01-01

    This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

  3. Sabores de viagem - as culinárias regionais nos guias turísticos entre 1966 e 1983.

    Directory of Open Access Journals (Sweden)

    Luciana Patrícia de Morais

    2012-11-01

    Full Text Available Partindo da premissa de que os saberes culinários se constituem em um dos principais suportes nos processos de construção e transmissão de identidades, busca-se entender a comida regional como expressão das culturas regionais em sua relação com a atividade turística. Parte-se do pressuposto de que a difusão dos regionalismos tem relação próxima com o desenvolvimento desta atividade no Brasil, num movimento em que as personalidades regionais são apropriadas e alçadas ao status de atrativos e produtos turísticos. Sem pretender estabelecer uma relação de causalidade entre turismo e comida regional, entende-se que embora a visibilidade das culinárias típicas no Brasil tenha relação com uma série de outras variáveis, o turismo é fundamental para se compreender os contornos e o alcance da visibilidade destas culinárias. Para tanto, a discussão terá como ponto de partida uma reflexão sucinta sobre os guias de viagem como fonte de pesquisa tomando como estudo de caso específico os casos de Minas Gerais e Paraná no interior do Guia Quatro Rodas Brasil entre 1966 e 1983.

  4. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre, E-mail: a.xhonneux@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Kasselmann, Stefan; Rütten, Hans-Jochem [Forschungszentrum Jülich, 52425 Jülich (Germany); Becker, Kai [Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Allelein, Hans-Josef [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany)

    2014-05-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  5. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    International Nuclear Information System (INIS)

    Xhonneux, Andre; Kasselmann, Stefan; Rütten, Hans-Jochem; Becker, Kai; Allelein, Hans-Josef

    2014-01-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  6. Commuters’ Exposure to Particulate Matter Air Pollution Is Affected by Mode of Transport, Fuel Type, and Route

    Science.gov (United States)

    Zuurbier, Moniek; Hoek, Gerard; Oldenwening, Marieke; Lenters, Virissa; Meliefste, Kees; van den Hazel, Peter; Brunekreef, Bert

    2010-01-01

    Background Commuters are exposed to high concentrations of air pollutants, but little quantitative information is currently available on differences in exposure between different modes of transport, routes, and fuel types. Objectives The aim of our study was to assess differences in commuters’ exposure to traffic-related air pollution related to transport mode, route, and fuel type. Methods We measured particle number counts (PNCs) and concentrations of PM2.5 (particulate matter ≤ 2.5 μm in aerodynamic diameter), PM10, and soot between June 2007 and June 2008 on 47 weekdays, from 0800 to 1000 hours, in diesel and electric buses, gasoline- and diesel-fueled cars, and along two bicycle routes with different traffic intensities in Arnhem, the Netherlands. In addition, each-day measurements were taken at an urban background location. Results We found that median PNC exposures were highest in diesel buses (38,500 particles/cm3) and for cyclists along the high-traffic intensity route (46,600 particles/cm3) and lowest in electric buses (29,200 particles/cm3). Median PM10 exposure was highest from diesel buses (47 μg/m3) and lowest along the high- and low-traffic bicycle routes (39 and 37 μg/m3). The median soot exposure was highest in gasoline-fueled cars (9.0 × 10−5/m), diesel cars (7.9 × 10−5/m), and diesel buses (7.4 × 10−5/m) and lowest along the low-traffic bicycle route (4.9 × 10−5/m). Because the minute ventilation (volume of air per minute) of cyclists, which we estimated from measured heart rates, was twice the minute ventilation of car and bus passengers, we calculated that the inhaled air pollution doses were highest for cyclists. With the exception of PM10, we found that inhaled air pollution doses were lowest for electric bus passengers. Conclusions Commuters’ rush hour exposures were significantly influenced by mode of transport, route, and fuel type. PMID:20185385

  7. Major results on the development of high density U-Mo fuel and pin-type fuel elements executed under the Russian RERTR program and in cooperation with ANL (USA)

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Stetsky, Y.; Suprun, V.; Dobrikova, I.; Trifonov, Y.; Mishunin, V.; Sorokin, V.

    2003-01-01

    VNIINM is active participant of 'Russian program on Reduced Enrichment for Research and Test Reactors'. Institute Works in two main directions: 1) development of new high-density fuels (HDF) and 2) development of new design of fuel elements with LEU. The development of the new type fuel element is carried out both for existing reactors, and for developing new advanced reactors. The 'TVEL' concern is coordinator of works of this program. The majority enterprises of branch (NIIAR, PIYaF, RRC KI, NZChK) take part in this work. Since 2000 these works are being conducted in cooperation with Argonne National Laboratory (USA) within the RERTR program under VNIINM with ANL contract. At the present, a large set of pre-pile investigations has been completed. All necessary fabrication procedures have been developed for utilization of U-Mo dispersion fuel in Russian-designed research reactors. For irradiation tests the pin-type mini-fuel elements with HDF dispersion fuel with LEU and the uranium density equaled to 4,0 and 6,0 g/cm 3 (up to 40 vol.%) have been manufactured. Their irradiation began in August 2003 in the MIR reactor (NIIAR, Dimitrovgrad). A large set of works for preparation of lifetime tests (WWR-M reactor in Gatchina) of two full-scale fuel assemblies with new pin-type fuel elements on basis LEU UO 2 -Al and UMo-Al fuels has been completed. The in-pile tests of fuel assemblies began in September 2003. The summary of important results of performed works and their near-term future are presented in paper. (author)

  8. Experimental study on DNB heat flux of plate-type fuel in pressurized condition

    International Nuclear Information System (INIS)

    Komori, Yoshihiro; Oshima, Kunio; Ishitsuka, Etsuo; Sakurai, Fumio; Sudo, Yukio; Saito, Minoru; Futamura, Yoshiaki; Kaminaga, Masanori.

    1992-07-01

    Experimental study was carried out in order to determine the DNB correlation for the safety analysis of the JMTR low enrichment fuel core. Since it is essential to examine applicability and safety margin of the correlation for the safety analysis, DNB heat fluxes were measured with the test section of rectangular flow channel simulating JMTR fuel element subchannel in the pressure range of 1 ∼ 13 kg/cm 2 abs and the velocity range of 0 ∼ 4.4 m/s. Reviewing existed DNB correlations based on the experimental data, Sudo correlations scheme was selected for the JMTR safety analysis with minor modification for the high flow rate region. Comparing the correlations scheme with experimental data, allowable limit of the minimum DNBR was determined to be 1.5. (author)

  9. Technical Appraisal of Continuous Destilator Type as Alternative Fuel Producer from Basic Materials of Arak Bali

    Directory of Open Access Journals (Sweden)

    Sukadana -

    2012-11-01

    Full Text Available Arrack Bali which is produced from traditional process has low quality (<40%. With controlling of operational variable such as evaporation temperature, will improve arrack Bali quality. Arrack Bali with quality more than 80 % has octane number more than 108,6, higher then petroleum octane number (80 until 90, easy burning and evaporation, very good to be alternative fuel to engine. In order to product height quality any operational variables like temperature, step, and sprayer models should be noticed. This experiment is to obtain operational variables of distillatory to product arrack Bali as an alternative fuel and it is tested in motor cycle engine at speed and compression ratio variables toward performance like emission. The higher evaporation temperature is the higher capacity of product to be obtained, on the other hand, the lower quality to be reached. Generally, comparing with petroleum, arrack Bali yields lower emission.

  10. Essential Data and Techniques for Conducting Microbial Fuel Cell and other Types of Bioelectrochemical System Experiments

    KAUST Repository

    Logan, Bruce E.

    2012-01-01

    Microbial fuel cells (MFCs) and other bioelectrochemical systems are new technologies that require expertise in a variety of technical areas, ranging from electrochemistry to biological wastewater treatment. There are certain data and critical information that should be included in every MFC study, such as specific surface area of the electrodes, solution conductivity, and power densities normalized to electrode surface area and volumes. Electrochemical techniques such as linear sweep voltammetry can be used to understand the performance of the MFC, but extremely slow scans are required for these biological systems compared to more traditional fuel cells. In this Minireview, the critical information needed for MFC studies is provided with examples of how results can be better conveyed through a full description of materials, the use of proper controls, and inclusion of a more complete electrochemical analysis. © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  12. Essential Data and Techniques for Conducting Microbial Fuel Cell and other Types of Bioelectrochemical System Experiments

    KAUST Repository

    Logan, Bruce E.

    2012-04-19

    Microbial fuel cells (MFCs) and other bioelectrochemical systems are new technologies that require expertise in a variety of technical areas, ranging from electrochemistry to biological wastewater treatment. There are certain data and critical information that should be included in every MFC study, such as specific surface area of the electrodes, solution conductivity, and power densities normalized to electrode surface area and volumes. Electrochemical techniques such as linear sweep voltammetry can be used to understand the performance of the MFC, but extremely slow scans are required for these biological systems compared to more traditional fuel cells. In this Minireview, the critical information needed for MFC studies is provided with examples of how results can be better conveyed through a full description of materials, the use of proper controls, and inclusion of a more complete electrochemical analysis. © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    Subbotin, V.; Alexeev, G.; Peskov, O.; Sapankevic, A.

    1976-01-01

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  14. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, V; Alexeev, G; Peskov, O; Sapankevic, A

    1976-08-01

    The conditions are formulated under which the results of the experimental research of the boiling. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented.

  15. Validation of Pressure Drop Models for PHWR-type Fuel Elements

    International Nuclear Information System (INIS)

    Brasnarof Daniel; Daverio, H.

    2003-01-01

    In the present work an one-dimensional pressure drop analytical model and the COBRA code, are validated with experimental data of CANDU and Atucha fuel bundles in low and high pressure experimental test loops.Models have very good agreement with the experimental data, having less than 5 % of discrepancy. The analytical model results were compared with COBRA code results, having small difference between them in a wide range of pressure, temperature and mass flow

  16. Development of proto-type advanced leaked fuel rod detection system

    International Nuclear Information System (INIS)

    Kang, Kyung Chul; Cho, Seong Won; Jeon, Jae Hyuk; Jeong, Jae Cheon; Kim, Min

    1996-02-01

    The fuel inspection equipment using ultrasonic signal has been developed its design and configuration in order to get inspection results more accurate and easier than the previous ones. In this task, the system functions are advanced by adopting of state of the art technologies in the field of digital servo control and signal processing. By the above endeavors, the total performance are improved and made to handle easily. 61 tabs., 31 figs., 3 ills., 9 refs. (Author)

  17. HCPWR type fuel elements design with respect to its control rods

    International Nuclear Information System (INIS)

    Abbate, M.J.; Sbaffoni, M.M.; Patino, N.E.; Torasso, O.

    1992-01-01

    The high conversion reactors (HCPWR) can improve the nuclear fuel utilization. One of its present problems is the optimization of the control rods' worth and its relationship with the void coefficient. This investigation means, starting from one reference's design, to analyze several possibilities on the number and distribution of the control rods. As main result, one design including 24 control rods in an optimized distribution, is recommended. (author)

  18. Physical properties, evaporation and combustion characteristics of nanofluid-type fuels

    OpenAIRE

    Tanvir, Saad

    2016-01-01

    Nanofluids are liquids with stable suspension of nanoparticles. Limited studies in the past have shown that both energetic and catalytic nanoparticles once mixed with traditional liquid fuels can be advantageous in combustion applications, e.g., increased energy density and shortened ignition delay. Contradictions in existing literature, scarcity of experimental data and lack of understanding on how the added nanoparticles affect the physical properties as well as combustion characteristics o...

  19. THE COMPLEX USE OF LOCAL TYPES OF FUEL IN THE POROUS CONSTRUCTION MATERIALS PRODUCTION

    Directory of Open Access Journals (Sweden)

    N. P. Voronova

    2014-01-01

    Full Text Available The article presents a comprehensive low-waste technology is the use of local fuels, which can be used in the technology of some porous building materials. Also provides new methods of preparation of porous building materials based on aggloporite using local fuels and waste energy on the basis of milled peat, fuel briquettes and wood chips allow to replace expensive imported components that comprise the raw mixtures (coal, anthracite.On the basis of mathematical modeling of cooling in reheat furnaces pusher drive developed a method of engineering calculation mode batch hardening in agglomeration. Submitted constructive solution for the development of the cooling charge with thermophysical rational justification cooling modes. A study of the temperature distribution within the charge depending on the different speeds of the belt sintering machine, and hence on the cooling time.The characteristics of the raw material deposits "Fanipol" and the optimal composition of the charge which includes loam, coal, milled peat. In industrial research obtained aggloporite this formulation has shown positive results in strength and density. Established that by decreasing the particle size of the fuel increases the redox potential of the combustion products, which reduces the height of the oxidizing zone and the speed of the sintering raw mix. These processes increase the productivity of sinter machine.Technology is implemented on the "Minsk factory of building materials". The tests analyzed production technology porous construction materials using milled peat with the addition of sawdust. The study results recommend further use of sapropel, which cost significantly lower raw material mixture of submissions and in their physical and mechanical properties much closer to the properties of milled peat.

  20. Analysis of steam explosions in plate-type, uranium-aluminum fuel test reactors

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1989-01-01

    The concern over steam explosions in nuclear reactors can be traced to prompt critical nuclear excursions in aluminum-clad/fueled test reactors, as well as to explosive events in aluminum, pulp, and paper industries. The Reactor Safety Study prompted an extensive analytical and experimental effort for over a decade. This has led to significant improvements in their understanding of the steam explosion issue for commercial light water reactors. However, little progress has been made toward applying the lessons learned from this effort to the understanding and modeling of steam explosion phenomena in aluminum-clad/fueled research and test reactors. The purposes of this paper are to (a) provide a preliminary analysis of the destructive events in test reactors, based on current understandings of steam explosions; (b) provide a proposed approach for determining the likelihood of a steam explosion event under scenarios in which molten U-Al fuel drops into a water-filled cavity; and (c) present a benchmarking study conducted to estimate peak pressure pulse magnitudes

  1. Realisations in the field of combustion for a new type of gaseous fuel based on hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Paunescu, L.; Surugiu, G. [Metallurgical Research Inst., Bucharest (Romania); Dica, C. [Rokura Industrial Applications, Bucharest (Romania); Stanescu, P.D. [Univ. for Technical Installation, Bucharest (Romania); Iorga, G. [Uzinsider Engineering, Galati (Romania); Necula, H. [Politechnica Univ., Bucharest (Romania); Ivan, I. [Mittal Steel, Galati (Romania)

    2006-07-01

    The trend towards the use of non-polluting energy sources to reduce or eliminate environmentally damaging combustion products such as carbon dioxide, carbon monoxide and nitrous oxides was discussed. Water electrolysis experiments were conducted to obtain an oxy-hydric gaseous fuel known as Klein gas. Klein gas contains hydrogen and oxygen in an almost stoichiometric proportion and has a unique molecular structure. From an energetic point of view, Klein gas behaves differently from other gases depending on the conditions where ignition and combustion occur. The temperature inside the flame varies from about 130 degrees C during free combustion under normal temperature and pressure conditions, up to the melting temperatures of some metals or refractory materials. If ignited Klein gas comes in contact with the surfaces of such a materials it can be used for cutting, brazing or welding. In order to use Klein gas in combustion installations such as industrial heating furnaces in iron and steel mills or in the ceramic and refractory industry, it should be used in combination with other gaseous fuels before ignition or by injection into an existing flame. This paper presented experimental results obtained by a Romanian team of researchers regarding the use of Klein gas in combustion installations with natural gas. The combustion rate was found to intensify as flame temperature increased, depending on the proportion of Klein gas used. The optimal proportion between the two fuels was found to be 1:5. 5 refs., 3 tabs., 4 figs.

  2. Selection of the new position for the instrumented fuel Flip type; Seleccion de la nueva posicion para el combustible instrumentado tipo Flip

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L C

    1991-12-15

    It was selected and proved the new position for the instrumented fuel, in function of the smallest influence on the temperature of the fuel due to movements of the control bars, as well as to its functionality regarding their geometric localization in the core. Therefore, it intends that the instrumented fuel Flip type, pass from the position C-9 to the C-6. It was selected the turn angle of the fuel of 180 C, for to be the one that presents a smaller influence of the control bars and it allows to register the values of temperature but high. It was as the movements of the control bars they produce local interferences of the neutron flow in the fuels and that in the case of the ring C, these interferences became maximum variations of the fuel temperature of 125 C, for operations to 1 MW in the thermal column. (Author)

  3. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    OpenAIRE

    H. P. Rahardjo; V. I. Sri Wardhani

    2017-01-01

    The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety an...

  4. Desempenho de diferentes guias de ondas para uso com o analisador de umidade TRASE Performance of different waveguides for use with the TRASE water content analyser

    Directory of Open Access Journals (Sweden)

    Eugênio F. Coelho

    2001-04-01

    Full Text Available O trabalho teve como objetivo avaliar a relação entre a umidade e a constante dielétrica aparente do solo e diferentes guias de onda para uso com o analisador de umidade Trase, que opera de acordo com o princípio da reflectometria no domínio do tempo - TDR. Amostras indeformadas e deformadas de duas manchas de textura diferente de um Latossolo Amarelo Distrófico foram retiradas do campo e acondicionadas em recipientes de 10 L, perfazendo quatro repetições para cada textura e estrutura. Foram construídas três diferentes guias de onda, com hastes de 0,15 m, sendo uma com capacitor e uma sem capacitor no início da guia, com espaçamento de 0,009 m entre hastes e outra sem capacitor, com espaçamento de 0,022 m entre hastes. Essas guias de onda, juntamente com guias originais do fabricante com hastes de 0,20 m, espaçadas 0,022 m, foram inseridas individualmente em cada recipiente. Dados de umidade obtidos gravimetricamente e pelo analisador TRASE, e da constante dielétrica, foram tomados usando-se todas as guias de onda em cada recipiente durante a secagem do solo, de 0,35 m³ m-3 a 0,10 m³ m-3. Três modelos matemáticos foram ajustados aos dados de umidade e da correspondente constante dielétrica do solo gerados pelas guias de onda do fabricante. Um modelo exponencial foi considerado como mais adequado para estimativas dos teores de água, em função da constante dielétrica para a guia de onda do fabricante. Todas as guias de onda avaliadas apresentam viabilidade de uso, desde que previamente calibradas.The objective of this work was to evaluate the relations between soil water content and the soil bulk dielectric constant, and to study different waveguides of a TRASE soil water content analyzer that operates according to TDR principles. Non-destructive and destructive samples of two sites of different texture of a Dystrophic Yellow Latossol were collected and packed into 10 L containers, resulting in four replications for each

  5. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  6. Sex-related differences in fuel utilization and hormonal response to exercise: implications for individuals with type 1 diabetes.

    Science.gov (United States)

    Brockman, Nicole K; Yardley, Jane E

    2018-06-01

    Sex-related differences in metabolic and neuroendocrine response to exercise in individuals without diabetes have been well established. Men and women differ in fuel selection during exercise, in which women rely to a greater extent on fat oxidation, whereas males rely mostly on carbohydrate oxidation for energy production. The difference in fuel selection appears to be mediated by sex-related differences in hormonal (including catecholamines, growth hormone, and estrogen) response to different types and intensities of exercise. In general, men exhibit an amplified counter-regulatory response to exercise, with elevated levels of catecholamines compared with women. However, women exhibit greater sensitivity to the lipolytic action of the catecholamines and deplete less of their glycogen stores than men during exercise, which suggests that women may experience a greater defense in blood glucose control after exercise than men. Conversely, little is known about sex-related differences in response to exercise in individuals with type 1 diabetes (T1D). A single study investigating sex-related differences in response to moderate aerobic exercise in individuals with T1D found sex-related differences in catecholamine response and fuel selection, but changes in blood glucose were not measured. To our knowledge, there are no studies investigating sex-related differences in blood glucose responses to different types and intensities of exercise in individuals with T1D. This review summarizes sex-related differences in exercise responses that could potentially impact blood glucose levels during exercise in individuals with T1D and highlights the need for further research.

  7. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  8. Quality verification for plate-type uranium-aluminum fuel elements for use in research reactors (Revision 1) - July 1976

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Paragraph (a) (7) of 50.34, Contents of Applications: Technical Information, of 10 CFR Part 50, Licensing of Production and Utilization Facilities, requires that each applicant for a construction permit to build a production or utilization facility include in its Preliminary Safety Analysis Report (PSAR) a description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the facility. The Regulatory Guide presented describes a method acceptable to the NRC staff for establishing and executing a quality assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research reactors

  9. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Kramer, J.M.

    1992-01-01

    The next step in the development of metal fuels for the integral fast reactor (IFR) is the conversion of the Experimental Breeder Reactor II (EBR-II) core to one containing the ternary U-20 Pu-10 Zr alloy clad with HT-9 cladding, i.e., the Mk-V core. This paper presents results of three hot-cell furnace simulation tests on irradiated Mk-V-type fuel elements (U-19 Pu-10 Zr/HT-9), which were performed to support the safety case for the Mk-V core. These tests were designed to envelop an umbrella (bounding) unlikely loss-of-flow (LOF) event in EBR-II during which the calculated peak cladding temperature would reach 776 degree C for < 2 min. The principal objectives of these tests were (a) demonstration of the safety margin of the fuel element, (b) investigation of cladding breaching behavior, and (c) provision of data for validation of the FPIN2 and LIFE-METAL codes

  10. Closed-form solution of a two-dimensional fuel temperature model for TRIGA-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rivard, J B [Sandia Laboratories (United States)

    1974-07-01

    If azimuthal power density variations are ignored, the steady-state temperature distribution within a TRIGA-type fuel element is given by the solution of the Poisson equation in two dimensions (r and z) . This paper presents a closed-form solution of this equation as a function of the axial and radial power density profiles, the conductivity of the U-ZrH, the inlet temperature, specific heat and flow rate of the coolant, and the overall heat transfer coefficient. The method begins with the development of a system of linear ordinary differential equations describing mass and energy balances in the fuel and coolant. From the solution of this system, an expression for the second derivative of the fuel temperature distribution in the axial (z) direction is found. Substitution of this expression into the Poisson equation for T(r,z) reduces it from a partial differential equation to an ordinary differential equation in r, which is subsequently solved in closed-form. The results of typical calculations using the model are presented. (author)

  11. SR 97: Post-closure safety for a KBS 3 type deep repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Hedin, A.; Kautsky, U.

    2000-03-01

    Prior to coming site investigations for siting of a deep repository for spent nuclear fuel, SKB has carried out the long-term safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion. The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. (author)

  12. Physical properties, evaporation and combustion characteristics of nanofluid-type fuels

    Science.gov (United States)

    Tanvir, Saad

    Nanofluids are liquids with stable suspension of nanoparticles. Limited studies in the past have shown that both energetic and catalytic nanoparticles once mixed with traditional liquid fuels can be advantageous in combustion applications, e.g., increased energy density and shortened ignition delay. Contradictions in existing literature, scarcity of experimental data and lack of understanding on how the added nanoparticles affect the physical properties as well as combustion characteristics of the resulting fuel motivated us to launch a detailed experimental and theoretical investigation. The surface tension of ethanol and n-decane based nanofluid fuels containing suspended nanoparticles were measured using the pendant drop method by solving the Young-Laplace equation. The results show that surface tension increases both with particle concentration (above a critical concentration) and particle size. This is because the Van der Waals forces between particles at the liquid/gas interface increases surface free energy that overcomes any electrostatic repulsion between the particles and increases surface tension. This present work also reports experimental analysis of the latent heat of vaporization ( Hfg) of nanofluids. Results show that the addition of Ag and Fe nanoparticles in water results is a substantial reduction in Hfg. On the contrary Al addition slightly increases Hfg. Similar observations are made for ethanol based nanofluids. Molecular dynamics simulations showed that the strength of bonding between particles and the fluid molecules is the governing factor in the variation of Hfg upon particle addition. The thermal conductivity was measured using KD2-Pro from Decagon Devices based on the transient line heat source method. The rheological properties of the ethanol and ethanol/nanoparticles suspensions are measured using a StresstechRTM rotational rheometer. Both properties increased with increasing particle concentration. Trends are found to be consistent

  13. Laboratory measurements of trace gas emissions from biomass burning of fuel types from the southeastern and southwestern United States

    Directory of Open Access Journals (Sweden)

    I. R. Burling

    2010-11-01

    Full Text Available Vegetation commonly managed by prescribed burning was collected from five southeastern and southwestern US military bases and burned under controlled conditions at the US Forest Service Fire Sciences Laboratory in Missoula, Montana. The smoke emissions were measured with a large suite of state-of-the-art instrumentation including an open-path Fourier transform infrared (OP-FTIR spectrometer for measurement of gas-phase species. The OP-FTIR detected and quantified 19 gas-phase species in these fires: CO2, CO, CH4, C2H2, C2H4, C3H6, HCHO, HCOOH, CH3OH, CH3COOH, furan, H2O, NO, NO2, HONO, NH3, HCN, HCl, and SO2. Emission factors for these species are presented for each vegetation type burned. Gas-phase nitrous acid (HONO, an important OH precursor, was detected in the smoke from all fires. The HONO emission factors ranged from 0.15 to 0.60 g kg−1 and were higher for the southeastern fuels. The fire-integrated molar emission ratios of HONO (relative to NOx ranged from approximately 0.03 to 0.20, with higher values also observed for the southeastern fuels. The majority of non-methane organic compound (NMOC emissions detected by OP-FTIR were oxygenated volatile organic compounds (OVOCs with the total identified OVOC emissions constituting 61 ± 12% of the total measured NMOC on a molar basis. These OVOC may undergo photolysis or further oxidation contributing to ozone formation. Elevated amounts of gas-phase HCl and SO2 were also detected during flaming combustion, with the amounts varying greatly depending on location and vegetation type. The fuels with the highest HCl emission factors were all located in the coastal regions, although HCl was also observed from fuels farther inland. Emission factors for HCl were generally higher for the southwestern fuels

  14. Characterization of char derived from various types of solid wastes from the standpoint of fuel recovery and pretreatment before landfilling

    International Nuclear Information System (INIS)

    Hwang, I.H.; Matsuto, T.; Tanaka, N.; Sasaki, Y.; Tanaami, K.

    2007-01-01

    Carbonization is a kind of pyrolysis process to produce char from organic materials under an inert atmosphere. In this work, chars derived from various solid wastes were characterized from the standpoint of fuel recovery and pretreatment of waste before landfilling. Sixteen kinds of municipal and industrial solid wastes such as residential combustible wastes, non-combustible wastes, bulky wastes, construction and demolition wastes, auto shredder residue, and sludges were carbonized at 500 deg. C for 1 h under nitrogen atmosphere. In order to evaluate the quality of char as fuel, proximate analysis and heating value were examined. The composition of raw waste had a significant influence on the quality of produced char. The higher the ratio of woody biomass in waste, the higher heating value of char produced. Moreover, an equation to estimate heating value of char was developed by using the weight fraction of fixed carbon and volatile matter in char. De-ashing and chlorine removal were performed to improve the quality of char. The pulverization and sieving method seems to be effective for separation of incombustibles such as metal rather than ash. Most char met a 0.5 wt% chlorine criterion for utilization as fuel in a shaft blast furnace after it was subjected to repeated water-washing. Carbonization could remove a considerable amount of organic matter from raw waste. In addition, the leaching of heavy metals such as chrome, cadmium, and lead appears to be significantly suppressed by carbonization regardless of the type of raw waste. From these results, carbonization could be considered as a pretreatment method for waste before landfilling, as well as for fuel recovery

  15. Evaluation of analysis method standardless by WDXRF and EDXRF of aluminum powder used in MTR type fuel

    International Nuclear Information System (INIS)

    Scapin, Valdirene O.; Salvador, Vera L.R.; Cotrim, Marycel E.B.; Pires, Maria A.F.; Scapin, Marcos A.

    2011-01-01

    The nuclear fuel used in IEA-R1m reactor at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP) is the MTR type. This fuel is compound of a core (U 3 Si 2 -Al dispersion briquette) wrapped in an aluminum plate with two cladding (superior and inferior) both in aluminum. The fuel element efficiency depends on the quality control of U 3 Si 2 and aluminum. For aluminum should be checked the impurities levels such as Si, Mn, Fe, Co, Cu, Zn and others and Al total . Aiming to provide a quick method, multielemental and non-destructive, the performance of the wavelength dispersive (WDXRF) and energy dispersive (EDXRF) X-ray fluorescence techniques, using the curve instrument sensitivity curve method, also known like standard less analysis, was evaluated. This method allows the determination from the element boron (Z=5) to uranium (Z=92) with concentrations ranging from 0.001 to 99.99% without the need for individual calibration curve and chemical pretreatments in the sample preparation. The results were compared with calibration curve method data, using statistical tests tools. By multivariate analysis of all the experimental data, especially by the discriminant analysis (DA) and cluster analysis (CA), respectively, it was possible to evaluate a correlation between variables of the applied analytical methods could be interpreted in context to qualify the fuels by XRF technique and method standard less. The results showed that the proposed method is satisfactory for both spectrometers; however it was found that the WDXRF presents the greatest conformity degree. (author)

  16. Competitiveness of peat briquets compared with other types of solid communal household fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dem' yanov, Ye S; Dayen, Ye I

    1979-01-01

    Approximately 20% solid, boiler furnace fuel, the greater part of which is coal, is presently used in communal-househould sector of the national economy of the country. However, the limited quantity of high grade coal, which leads to deliveries of coal with an increased fines content to the communal and domestic sector, as well as a significant sulfur content in it (3-6%) justified an examination of the conditions in which the portion of the demand for communal and domestic needs can be increased, primarily in the European part of the USSR, in the peat briquets. Both the low content of sulfur in them (approximately 0.3%), as well as the high useful utilization factor of the briquets with burning (approximately 81.7%) speak in favor of an expansion in the use of peat briquets. An important feature in increasing the competitiveness of the peat briquets is a reduction in the cost of their production. So that the peat briquets could compare in realization conditions with, for instance, Donetsk coal, the coast of 1 t of briquets at peat briquet factories must not exceed 11 ruble, 32 kopeks. The Moscow branch of VNIITP is developing an economic and mathematical model of the distribution of solid fuel to communal domestic needs in a region of the country. The model is designed for computer calculation of the optimal variants of providing the population with fuel which makes it possible to reduce expenditures on transport, to reduce the costs of domestic organizations and to increase profits, depending on the optimization criterion used.

  17. Re-irradiation and limit testing of the fuels PWR type reactors

    International Nuclear Information System (INIS)

    Roche, M.; Molvault, M.

    1978-01-01

    In view of investigating the neutron radiation behavior of PWR fuel pins, the S.P.S. (Services des Piles de Saclay) developed a set of experimental means used at OSIRIS in Saclay Nuclear Research Center. Said devices are shown to be able to meet present problems concerning can failures, power and temperature cycling, remote-control studies. These means can also be used either for statistical studies, they can then receive several samples, or for analytical studies in instrumented devices of large capacity and accelerated irradiation rate [fr

  18. Numerical Modeling of Limiting Oxygen Index Apparatus for Film Type Fuels

    Directory of Open Access Journals (Sweden)

    Amit Kumar

    2012-12-01

    Full Text Available A detailed three-dimensional numerical model is used to compute the flow pattern and the flame behavior of thin solid fuels in a rectangular column that resembles a standard Limiting Oxygen Index (LOI device. The model includes full Navier-Stokes equations for mixed buoyant-forced flow and finite rate combustion and pyrolysis reactions so that the sample LOI can be computed to study the effect of feeding flow rate, sample width and gravity levels. In addition to the above parameters, the sample location in the column and the column cross-sectional area are also investigated on their effect on the ambient air entrainment from the top.

  19. Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel; Evaluacion Analitica de 4 Metodos de Determinacion de PAHs medianteHPLC en un Fuel de Tipo II

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

    2009-05-21

    An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

  20. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    International Nuclear Information System (INIS)

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

    2006-01-01

    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  1. Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

    International Nuclear Information System (INIS)

    Durazzo, M.

    1985-01-01

    The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

  2. Safety criterion for burnout of the plate-type fuel in pressurized conditions

    International Nuclear Information System (INIS)

    Komori, Y.; Kaminaga, M.; Sakurai, F.; Ando, H.; Sudo, Y.; Saito, M.; Futamura, Y.

    1992-01-01

    The reduced enrichment program for JMTR is now underway and the core conversion to LEU (Low Enrichment Uranium) is scheduled to be made in 1993. Consistent with the safety guide which have been recently developed for research and test reactors in Japan, the safety analysis for the JMTR LEU conversion was conducted. In the safety analysis, DNB (Departure from Nucleate Boiling) heat flux correlation for the JMTR downflow condition was reconsidered because recent studies on burnout show that DNB heat fluxes with thin rectangular channels under low flow rate and low pressure conditions are much lower than predicted values by conventional DNB correlations. Available DNB data, however, are very limited for the JMTR operation pressure range, so that DNB experiments were conducted simulating the JMTR fuel subchannel. Based mainly on the present experimental data, the DNB correlations scheme composed of three correlations was selected for the JMTR safety analysis. Errors of the correlations scheme with experimental data were evaluated in order to determine the allowable limit of the minimum DNB ratio for preventing fuel failure. (author)

  3. CASTOR(r) and CONSTOR(r) type transport and storage casks for spent fuel and high active waste

    International Nuclear Information System (INIS)

    Kuehne, B.; Sowa, W.

    2002-01-01

    The German company GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and high-level waste. CASTOR(r) casks are used at 18 sites on three continents. Spent fuel assemblies of the types PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste (HAW) containers are stored in these kinds of casks. More than 600 CASTOR(r) casks have been loaded for long-term storage. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in each case. There is no indication that problems will arise in the future. Of course, the experience of 20 years has resulted in improvements of the cask design. One basic improvement is GNB's development since the mid 1990s of a sandwich cask design using heavy concrete and steel as basic materials, for economical and technical reasons. This CONSTOR(r) cask concept also fulfils all design criteria for transport and storage given by the IAEA recommendations and national authorities. By May 2002 40 CONSTOR(r) casks had been delivered and 15 had been successfully loaded and stored. In this paper the different types of casks are presented. Experiences gained during the large number of cask loadings and more than 4000 cask-years of storage will be summarised. The presentation of recent and future development shows the optimisation potential of the CASTOR(r) and CONSTOR(r) cask families for safe and economical management of spent fuel. (author)

  4. Report on research outline in 1999 fiscal year on public collection type research of nuclear fuel cycle.

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The Japan Nuclear Cycle Development Institute (JNC) initiated a system to promote the public collection type researches of nuclear fuel cycle since 1999, in order to intend to promote collaboration with universities and research institutes. The system aims to intend to activate researching environment in JNC by intercourse, information exchange, publication of results and so on between researchers in and out of JNC, effectively to proceed basic and fundamental R and D. Here were outlined on eleven researching theme, such as upgrading of velocity and temperature measurement of non-contact type fluid by using electromagnetic ultrasonic wave, development of temperature history memory elements for radiation environment test of a fast breeder reactor, direct glassification of salt wastes by using oxygen plasma, a method on dissolving extraction of metal oxides by using a super-critical fluid, creation of highly separable materials by assistance of computational chemistry (aiming at high order processing of high level radioactive wastes), and so on. (G.K.)

  5. Report on research outline in 1999 fiscal year on public collection type research of nuclear fuel cycle

    International Nuclear Information System (INIS)

    2000-07-01

    The Japan Nuclear Cycle Development Institute (JNC) initiated a system to promote the public collection type researches of nuclear fuel cycle since 1999, in order to intend to promote collaboration with universities and research institutes. The system aims to intend to activate researching environment in JNC by intercourse, information exchange, publication of results and so on between researchers in and out of JNC, effectively to proceed basic and fundamental R and D. Here were outlined on eleven researching theme, such as upgrading of velocity and temperature measurement of non-contact type fluid by using electromagnetic ultrasonic wave, development of temperature history memory elements for radiation environment test of a fast breeder reactor, direct glassification of salt wastes by using oxygen plasma, a method on dissolving extraction of metal oxides by using a super-critical fluid, creation of highly separable materials by assistance of computational chemistry (aiming at high order processing of high level radioactive wastes), and so on. (G.K.)

  6. Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Andrzejewski, Claudio de Sa

    2005-07-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  7. Avaliação das guias condilar e incisal em função da curva de compensação e da altura das cúspides - Releitura das Leis de Articulação de Hanau

    OpenAIRE

    Roger Nishyama

    2011-01-01

    Este estudo teve como objetivo avaliar a variação das guias sagitais (guia condilar e guia incisal), em função do plano de orientação (curvatura e inclinação) e da altura das cúspides, no movimento protrusivo mandibular, in vitro, buscando as relações entre esses fatores, tanto de forma analógica como digital. No estudo analógico foram avaliadas as inclinações das guias condilares, inclinações da guia incisal, proeminência da curva de compensação e inclinação do plano oclusal. Para isso foram...

  8. Guide for the development of power generation projects using renewable energy in and for municipalities; Guia para el desarrollo de proyectos de generacion de electricidad con energia renovable en y para los municipios

    Energy Technology Data Exchange (ETDEWEB)

    De Buen Rodriguez, Odon [Energia, Tecnologia y Educacion, S.C., ENTE, S.C. (Mexico)

    2010-05-15

    The Mexico Competitiveness Project has come about as a result of a good understanding and broad cooperation between the United States and Mexican governments. Its primary objective is to strengthen governance so as to improve the economic competitiveness of Mexico as well as to improve environmental management for the preservation of biodiversity in the country. This program, financed by the United States Agency for International Development (USAID), will be carried out for approximately 4 years, from October 2008 to November 2012. Its cooperative efforts are primarily focused on assisting local efforts to foster sustainable development. In this context, and as part of the specific activities on which the program is focused during the current year, the Guide for the Development of Power Generation Projects in Municipalities was designed, whose purpose is to integrate information to enable municipal officials to understand these types of projects and take actions to foster their development. The present guide consists of a set of modules that detail and explain the most important aspects involved in the development of these projects, ranging from legal considerations to costs associated with the projects themselves. This guide has two sections: a general portion with 8 chapters about topics directly related to the projects and a set of appendices that describe aspects relevant to moving forward with projects for power generation using renewable energy. [Spanish] El Programa de Competitividad Mexico se deriva del buen entendimiento y amplia cooperacion entre los gobiernos de Estados Unidos y Mexico, y tiene como principal objetivo fortalecer la gobernabilidad para mejorar la competitividad economica de Mexico, asi como robustecer la gestion ambiental para preservar la biodiversidad del pais. Este programa, que es financiado por la Agencia de los Estados Unidos para el Desarrollo Internacional (USAID), tiene una duracion aproximada de cuatro anos a partir de octubre

  9. Transmutation of long-lived nuclides in the fuel cycle of Brest-type reactors

    International Nuclear Information System (INIS)

    Lopatkin, A.V.; Orlov, V.V.; Filin, A.I.

    2001-01-01

    Transmutation of long-lived nuclides produced as a result of nuclear generation, should be set up proceeding from the principle of reasonable sufficiency, expressed as radiation equivalence between the radwaste sent to disposal and source natural uranium. In this case, introduction of fast reactors of new generation (such as BREST or other reactors based on similar philosophy) will resolve transmutation problems even with the thermal-to-fast reactor capacity ratio of 2:1. The authors of the 'Strategy of nuclear power development in Russia' foresee, and substantiate their prediction, that fast reactors of the new generation will account for no less than 2/3 of nuclear capacity in future large-scale nuclear power sector. Fast reactors will be the basis of a transmutation fuel cycle, which will remove the need of creating additional transmutation facilities. (author)

  10. RA-3 reactor core with uranium silicide fuel elements P-07 type

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2003-01-01

    Following the studies on the utilization of fuel elements (FE) containing uranium silicide, core of the RA-3 was analyzed with several calculation models. At first, the present situation, i.e. the core charged with normal FE (U 3 O 8 ), has been analyzed to validate the simulation methodology comparing with experimental results and to establish reference data to 5 and 10 MW able to be compared with future new situations. Also, CITVAP's nuclear data libraries to be used in irradiation experiment planning were completed. The results were satisfactory and were applied to the study of the core containing P-07 FE [U 3 Si 2 ], in face of a future core change. Comparing with the performance of the U 3 O 8 FE, the silicides ones show the following advantages: - average burnup: 45 % greater; -extraction burnup increase 12 %; and, -the residence time [in full power days] could be a 117 % greater. (author)

  11. Performance of a novel type of electrolyte-supported solid oxide fuel cell with honeycomb structure

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz-Morales, Juan Carlos; Savvin, Stanislav N.; Nunez, Pedro [Departmento de Quimica Inorganica, Universidad de La Laguna, 38200 Tenerife (Spain); Marrero-Lopez, David [Departamento de Fisica Aplicada I, Universidad de Malaga, 29071 Malaga (Spain); Pena-Martinez, Juan; Canales-Vazquez, Jesus [Instituto de Energias Renovables-Universidad de Castilla la Mancha, 02006 Albacete (Spain); Roa, Joan Josep; Segarra, Merce [DIOPMA, Departamento de Ciencia de los Materiales e Ing. Metalurgica, 08028 Barcelona (Spain)

    2010-01-15

    A novel design, alternative to the conventional electrolyte-supported solid oxide fuel cell (SOFC) is presented. In this new design, a honeycomb-electrolyte is fabricated from hexagonal cells, providing high mechanical strength to the whole structure and supporting the thin layer used as electrolyte of a SOFC. This new design allows a reduction of {proportional_to}70% of the electrolyte material and it renders modest performances over 320 mW cm{sup -2} but high volumetric power densities, i.e. 1.22 W cm{sup -3} under pure CH{sub 4} at 900 C, with a high OCV of 1.13 V, using the standard Ni-YSZ cermet as anode, Pt as cathode material and air as the oxidant gas. (author)

  12. Experimental research on velocity distribution in narrow slots of plane type reactor fuel

    International Nuclear Information System (INIS)

    Qu Xinxing; Zhang Youjie; Jia Haijun; Jiang Shengyao; Bo Hanliang; Min Gang

    2003-01-01

    The experimental research on velocity distribution in multiple parallel narrow channels formed by fuel plane of assembly is carried out under various Re and the water without ions as fluid in testing loop. The experimental results show that under various Re within a channel the velocity in the middle area is high and the velocity in the edge of channel decreases rapidly to zero, and the velocity is symmetrically distributed along the central line, thus the velocity distribution within a channel is like a trapezium. While in parallel channels the velocity of the middle channel is low and the velocity of the channel close to the edge is high, it is supposed to be caused by the inlet structure of channels, this concave distribution is a disadvantage to the even distribution of reactor flux

  13. Development of advanced types of fuel at JSC 'UMP'

    International Nuclear Information System (INIS)

    Gagarin, A.E.; Kuchkovskiy, A.; Bezhetskiy, S.V.; Vachnenko, V.V.; Manych, A.V.; Russin, Y.G.

    2009-01-01

    As compared with all acting today techniques of power generation only nuclear power engineering has the real reserve of fuel and contaminate the environment minimum. As IAEA expert assessments show, up to 130 new power generating units with total power 430 gigaWt and annual energy production up to 3000 billion kilowatt-hours will be built in the world to 2020 year. It can be up to 30% of the world power balance. Kazakhstan today has unprecedented positions for nuclear power engineering development in the Republic as well as possibility to have an influence on its development in the world. Actually for today National Atomic Company (NAC) 'Kazatomprom' has the 4th position in the world in satisfaction of world demand for uranium, but to 2010 Kazakhstan should be the world leader in uranium mining in accordance with the company strategy

  14. Dose calculation for accident situations at WWR-S type spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Margeanu, S.; Florescu, G.

    2006-01-01

    Full text: The Spent Nuclear Fuel Repository at IFIN-HH Bucharest (SNFR IFIN-HH) consists in four pools, repository hall, radiological monitoring system, ventilation system and auxiliary systems. At the moment the remaining activity in the repository is about 3500 Ci. Despite of the small activity, for emergency preparedness purposes, several accident scenarios, with a non zero probability of occurrence during the repository lifetime, have been postulated. Evaluations of radiological consequences to personnel, general public and environment, for each accident scenario have been performed. The radioactive inventory was evaluated with ORIGEN code from SCALE computer code system and radiological consequences were evaluated with COSYMA computer code. Assumptions for the source term determination, meteorological conditions and release, are presented. The calculated values of doses and risk are also presented. The impact of these accident scenarios on population and environment is also discussed. (authors)

  15. Efficiency of SOFC type fuel cells; Eficiencia de celulas combustiveis do tipo SOFC

    Energy Technology Data Exchange (ETDEWEB)

    Amaral, Alexandre Alves do; Matos, Francisco F.; Boaventura, Jaime S.; Benedicto, Joao Paulo S.; Alencar, Marcelo [Universidade Federal da Bahia (UFBA), Salvador, BA (Brazil). Inst. de Quimica. Dept. de Fisico-Quimica

    2006-07-01

    production of electric energy, mainly due to the fact that it is virtually no pollutant. Typically, the SOFC is constituted of at least seven distinct phases: fuel, anode, electrolyte substrate (separating the two electrodes), cathode, air, and electrical inter connectors (completing the electrical circuit). Thermodynamics clearly shows that electrochemical systems only can be reversible when homogeneous, what it is not case of the Sofc. Therefore, the application of equilibrium thermodynamics to these systems is incorrect. This work proposes that the Sofc can be better depicted from reactions between adsorbed species. The efficiency then is calculated as the ratio between the free energy of these reactions to the combustion heat. Thermodynamic parameter estimative is developed for the global and adsorbed species reactions. (author)

  16. The theoretical possibility of reducing the doubling time in a fast-reactor by using heterogeneous configurations of various types of fuel

    International Nuclear Information System (INIS)

    Orlov, V.V.; Slesarev, I.S.; Zaritskij, S.M.; Subbotin, S.A.; Alekseev, P.N.; Zverkov, Yu.A.

    1980-01-01

    The authors have derived approximate expressions relating the doubling time of a fast reactor using various types of fuel simultaneously to the doubling time of traditional (homogeneous) reactors in which these types of fuel are used separately. These relationships afford a means of determining the conditions in which the use of various types of fuel can result in an improved doubling time. It was established that the use of heterogeneous compositions formed from assemblies of homogeneous systems gives a notable gain in doubling time over that of any of the original homogeneous systems if the doubling times were similar to each other. This gain is fairly large even in the case of BN reactors with high fuel volume fractions. The size of the gain depends on the degree of ''differentiation'' in the neutron and thermal properties of the components of the heterogeneous reactor. An optimum proportion has been found for the assemblies taken from the original homogeneous systems, governed primarily by the ratio of fuel densities. Estimates were made of the advantages of metallic oxide compositions over the traditional compositions used in large, fast reactors of the BN type. These estimates indicate that the former can be considered as alternative homogeneous compositions with carbide or nitride fuel as far as breeding characteristics are concerned. (author)

  17. Dilatational behaviour of ZrNb1 fuel cans of a WWER-type reactor during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Stephan, M.; Wetzel, L.

    1987-01-01

    Based on an assessment of various factors of influence on the performance of fuel cans during normal operation and imaginable accidents, the necessity of studying creep and burst behaviour of WWER-type fuel cans of ZrNb1 under simulated LOCA conditions has been proved and an experimental facility designed for this purpose is described. Control of fuel can temperature is accomplished through a minicomputer during the creep and bursts experiments. With this, various temperature loading profiles of the fuel cans can be realized. Experimental results on dilatational behaviour of ZrNb1 fuel cans from isothermal creep and burst experiments in air are presented and compared with values for Zircaloy. (author)

  18. Development of fabrication process of upper nozzle BIBLIS type of PWR fuel element

    International Nuclear Information System (INIS)

    Miranda, O.; Lorenzo, D.F.R.

    1982-01-01

    Process and parameters of milling and welding of a upper nozzle BIBLIS type prototype are presented. Milling process, cutting tools studies, production devices and inspection were developed and researched. (author) [pt

  19. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  20. Evaluation of Erosion of the Dummy ''EE'' Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    International Nuclear Information System (INIS)

    Brower, Jeffrey O.; Glazoff, Michael V.; Eiden, Thomas J.; Rezvoi, Aleksey V.

    2016-01-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and ''horseshoeing'' defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum ''dummy'' plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth ''S'' curve, was represented by a series temperature rise ''humps,'' which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping and

  1. Um guia de recomendações para o emprego de dados multimídia em sites web

    OpenAIRE

    Correa, Rodrigo Stéfani

    2002-01-01

    Dissertação (mestrado) - Universidade Federal de Santa Catarina, Centro Tecnológico. Programa de Pós-Graduação em Ciência da Computação. Com a evolução do suporte tecnológico da Web/Internet, surgiram diversos sites web multimídia. Atualmente existem poucos trabalhos voltados à orientação dos desenvolvedores quanto ao uso adequado dos dados multimídia. Este trabalho propõe um guia de recomendações para o emprego de dados multimídia para auxiliar o projeto de sites web. Para alcançar os obj...

  2. Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

    Science.gov (United States)

    Robel, Martin; Kristo, Michael J

    2008-11-01

    The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

  3. Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

    International Nuclear Information System (INIS)

    Rossa, Riccardo; Borella, Alessandro; Labeau, Pierre-Etienne; Pauly, Nicolas; Meer, Klaas van der

    2015-01-01

    The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of 239 Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of 239 Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a 239 Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to 239 Pu, in comparison with a 235 U fission chamber, with a 3 He proportional counter, and with a 10 B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the 239 Pu and 235 U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the 3 He and 10 B proportional counters to increase the sensitivity to 239 Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies

  4. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type

    International Nuclear Information System (INIS)

    Lucatero, M.A.; Hernandez L, H.

    2003-01-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  5. Guia de atencion integral para la prevencion, deteccion temprana y tratamiento de las complicaciones del embarazo: Sección de Toxoplasmosis

    Directory of Open Access Journals (Sweden)

    Jorge Alberto Cortes

    Full Text Available Resumen Como parte de la Guia de atencion integral para la prevencion, deteccion temprana y tratamiento de las complicaciones del embarazo, desarrollada y financiada por el Departamento de Ciencia, Tecnologia e Innovacion, Colciencias, y el Ministerio de Salud y Proteccion Social de Colombia, se seleccionó a la infeccion por Toxoplasma para el desarrollo de recomendaciones para su prevencion, diagnostico y tratamiento. La infeccion por Toxoplasma gondii (T. gondii durante el embarazo puede resultar en graves complicaciones para el feto y dejar importantes secuelas al recien nacido. Se realizo una guia basada en la mejor evidencia disponible en la literatura cientifica, con especial pertinencia a la informacion colombiana. Un consenso de expertos en parasitologia, ginecologia, neonatologia e infectologia, tanto de adultos como pediatrica, desarrollo las recomendaciones. Se propone que las recomendaciones de esta guia de atencion integral sean utilizadas por los profesionales de salud de los programas de atencion del embarazo del pais con el fin de disminuir la morbilidad y mortalidad atribuible a esta enfermedad. Se formulan recomendaciones especificas para el diagnostico desde el primer trimestre, consejos de prevencion en las mujeres no infectadas, identificacion de la infeccion del feto o del recien nacido y recomendaciones de tratamiento en estos escenarios.

  6. Flashback Characteristics of Syngas-Type Fuels Under Steady and Pulsating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tim Lieuwen

    2007-09-30

    The objective of this project was to improve understanding and modeling of flashback, a significant issue in low emissions combustors containing high levels of hydrogen. Experimental studies were performed over a range of fuel compositions, flow velocities, reactant temperatures, and combustor pressures to study the factors leading to flashback. In addition, high speed imaging of the flashback phenomenon was obtained. One of the key conclusions of this study was that there existed multiple mechanisms which lead to flashback, each with different underlying parametric dependencies. Specifically, two mechanisms of 'flashback' were noted: rapid flashback into the premixer, presumably through the boundary layer, and movement of the static flame position upstream along the centerbody. The former and latter mechanisms were observed at high and low hydrogen concentrations. In the latter mechanism, flame temperature ratio, not flame speed, appeared to be the key parameter describing flashback tendencies. We suggested that this was due to an alteration of the vortex breakdown location by the adverse pressure gradient upstream of the flame, similar to the mechanism proposed by Sattelmayer and co-workers [1]. As such, a key conclusion here was that classical flashback scalings derived from, e.g., Bunsen flames, were not relevant for some parameter regimes found in swirling flames. In addition, it was found that in certain situations, pure H2 flames could not be stabilized, i.e., the flame would either flashback or blowout at ignition. This result could have significant implications on the development of future high hydrogen turbine systems.

  7. Temperature modeling for analysis and design of the sintering furnance in HTR fuel type of ball

    International Nuclear Information System (INIS)

    Saragi, Elfrida; Setiadji, Moch

    2013-01-01

    One of the factors that determine the safety of the operation of the sintering furnace fuel HTR ball is the temperature distribution in the ceramic tube furnace. The temperature distribution must be determined at design stage. The tube has a temperature of 1600 °C at one end and about 40 °C at the other end. The outside of the tube was cooled by air through natural convection. The tube is a furnace ceramic tube which its geometry are 0.08, 0.09 and 0.5 m correspondingly for the inner tube diameter, outer tube diameter and tube length. The temperature distribution of the tube is determined by the natural convection coefficient (NCF), which is difficult to be calculated manually. The determination of NCF includes the Grasshoff, Prandtl, and Nusselt numbers which is a function of the temperature difference between the surrounding air with the ceramic tube. If the temperature vary along the tube, the complexity of the calculations increases. Thus the proposed modeling was performed to determine the temperature distribution along the tube and heat transfer coefficient using a self-developed software which permit the design process easier

  8. A nodal method of calculating power distributions for LWR-type reactors with square fuel lattices

    International Nuclear Information System (INIS)

    Hoeglund, Randolph.

    1980-06-01

    A nodal model is developed for calculating the power distribution in the core of a light water reactor with a square fuel lattice. The reactor core is divided into a number of more or less cubic nodes and a nodal coupling equation, which gives the thermal power density in one node as a function of the power densities in the neighbour nodes, is derived from the neutron diffusion equations for two energy groups. The three-dimensional power distribution can be computed iteratively using this coupling equation, for example following the point Jacobi, the Gauss-Seidel or the point successive overrelaxation scheme. The method has been included as the neutronic model in a reactor core simulation computer code BOREAS, where it is combined with a thermal-hydraulic model in order to make a simultaneous computation of the interdependent power and void distributions in a boiling water reactor possible. Also described in this report are a method for temporary one-dimensional iteration developed in order to accelerate the iterative solution of the problem and the Haling principle which is widely used in the planning of reloading operations for BWR reactors. (author)

  9. Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

    1979-05-01

    The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

  10. Variation of power generation at different buffer types and conductivities in single chamber microbial fuel cells

    KAUST Repository

    Nam, Joo-Youn

    2010-01-15

    Microbial fuel cells (MFCs) are operated with solutions containing various chemical species required for the growth of electrochemically active microorganisms including nutrients and vitamins, substrates, and chemical buffers. Many different buffers are used in laboratory media, but the effects of these buffers and their inherent electrolyte conductivities have not been examined relative to current generation in MFCs. We investigated the effect of several common buffers (phosphate, MES, HEPES, and PIPES) on power production in single chambered MFCs compared to a non-buffered control. At the same concentrations the buffers produced different solution conductivities which resulted in different ohmic resistances and power densities. Increasing the solution conductivities to the same values using NaCl produced comparable power densities for all buffers. Very large increases in conductivity resulted in a rapid voltage drop at high current densities. Our results suggest that solution conductivity at a specific pH for each buffer is more important in MFC studies than the buffer itself given relatively constant pH conditions. Based on our analysis of internal resistance and a set neutral pH, phosphate and PIPES are the most useful buffers of those examined here because pH was maintained close to the pKa of the buffer, maximizing the ability of the buffer to contribute to increase current generation at high power densities. © 2009 Elsevier B.V. All rights reserved.

  11. Variation of power generation at different buffer types and conductivities in single chamber microbial fuel cells.

    Science.gov (United States)

    Nam, Joo-Youn; Kim, Hyun-Woo; Lim, Kyeong-Ho; Shin, Hang-Sik; Logan, Bruce E

    2010-01-15

    Microbial fuel cells (MFCs) are operated with solutions containing various chemical species required for the growth of electrochemically active microorganisms including nutrients and vitamins, substrates, and chemical buffers. Many different buffers are used in laboratory media, but the effects of these buffers and their inherent electrolyte conductivities have not been examined relative to current generation in MFCs. We investigated the effect of several common buffers (phosphate, MES, HEPES, and PIPES) on power production in single chambered MFCs compared to a non-buffered control. At the same concentrations the buffers produced different solution conductivities which resulted in different ohmic resistances and power densities. Increasing the solution conductivities to the same values using NaCl produced comparable power densities for all buffers. Very large increases in conductivity resulted in a rapid voltage drop at high current densities. Our results suggest that solution conductivity at a specific pH for each buffer is more important in MFC studies than the buffer itself given relatively constant pH conditions. Based on our analysis of internal resistance and a set neutral pH, phosphate and PIPES are the most useful buffers of those examined here because pH was maintained close to the pK(a) of the buffer, maximizing the ability of the buffer to contribute to increase current generation at high power densities. Copyright 2009 Elsevier B.V. All rights reserved.

  12. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

  13. Study of homogeneous fuel cells type 10 x 10; Estudio de celdas de combustible homogeneas tipo 10 x 10

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Perusquia, R.; Ortiz, J.J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Francois, J.L.; Marquez, C.M. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2005-07-01

    At the moment in the National Institute of Nuclear Research (ININ) are carried out studies with the purpose of to establish a methodology that allows to carry out the neutron design of fuel cells of type 10 x 10. During the initial stage of the process of cells design, starting from the data that have to do with the planned energy demand it requires to be estimated the average value of the enrichment in U{sup 235} w/o of the one assemble. The experience has shown that the accuracy that is achieved in this estimate it depends, among other factors, of the information (e.g. concentrations of U{sup 235} and Gd{sub 2}O{sub 3}) of the cells that its are disposed in that moment. For what we consider convenient to enlarge the available information by means of a series of calculations of cell physics; and to the one same time some aspects can be studied on the parameters that define the characteristics of a fuel cell. In this work the effect of the presence of different distributions of the concentrations of the fissile material is analyzed and of burnup poisons on the reactivity parameters of the cell as well as in the peak factor of local power (LPPF-Local Power Peaking Factor). (Author)

  14. Factors affecting the performance of a single-chamber microbial fuel cell-type biological oxygen demand sensor.

    Science.gov (United States)

    Yang, Gai-Xiu; Sun, Yong-Ming; Kong, Xiao-Ying; Zhen, Feng; Li, Ying; Li, Lian-Hua; Lei, Ting-Zhou; Yuan, Zhen-Hong; Chen, Guan-Yi

    2013-01-01

    Microbial fuel cells (MFCs) are devices that exploit microorganisms as biocatalysts to degrade organic matter or sludge present in wastewater (WW), and thereby generate electricity. We developed a simple, low-cost single-chamber microbial fuel cell (SCMFC)-type biochemical oxygen demand (BOD) sensor using carbon felt (anode) and activated sludge, and demonstrated its feasibility in the construction of a real-time BOD measurement system. Further, the effects of anodic pH and organic concentration on SCMFC performance were examined, and the correlation between BOD concentration and its response time was analyzed. Our results demonstrated that the SCMFC exhibited a stable voltage after 132 min following the addition of synthetic WW (BOD concentration: 200 mg/L). Notably, the response signal increased with an increase in BOD concentration (range: 5-200 mg/L) and was found to be directly proportional to the substrate concentration. However, at higher BOD concentrations (>120 mg/L) the response signal remained unaltered. Furthermore, we optimized the SCMFC using synthetic WW, and tested it with real WW. Upon feeding real WW, the BOD values exhibited a standard deviation from 2.08 to 8.3% when compared to the standard BOD5 method, thus demonstrating the practical applicability of the developed system to real treatment effluents.

  15. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  16. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  17. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

  18. Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Medvedev, Pavel G.

    2009-01-01

    Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

  19. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

  20. Mechanical behaviour and failure of fuel cladding zirconium alloys in nuclear power plants under accidental RIA-type situation

    International Nuclear Information System (INIS)

    Doan, D.T.

    2009-01-01

    In French Nuclear Pressurized Water Reactors (PWRs), most of structural parts of the fuel assembly consist of zirconium alloy tubes and plates. Optimizing the management of fuel in nuclear power plants led to the increase in the duration of fuel cycles and power. The use of high fuel burnups requires drastic changes in the rules for reactor design in the nuclear safety. The evaluation of nuclear reactors in accident situations is based on reference accident scenarios. One of these hypothetical accidents, examined in this study, is the 'Reactivity Initiated Accident'. In order to assess the structural integrity of these parts it is necessary to characterize both the plastic flow and fracture behaviour of the materials at various stages of the life cycle, (i.e. at increasing levels of hydriding, irradiation, oxidation or thermal mechanical loading). The purpose of this work is to provide experimental data and to develop a model of the thermo-mechanical behaviour and to propose a design analysis method in the case of non-irradiated clads, in RIA-type situations. Mechanical tests were conducted on Cold-Worked-Stress-Relieved and on Recrystallized Zircaloy-4 sheets using various kinds of samples including smooth and notched tensile specimens and small punch tests. Temperature was set to 25, 250 and 600 C with hydrogen contents between 0 and 1000 ppm. The model is based on a simplified description of a Zircaloy polycrystal in which scalar isotropic ductile damage including void nucleation and growth is added. The model is also physically based to easily transfer parameters determined for one material state to another (e.g. transfer between sheet and tube or between different levels of irradiation). The model was implemented in the Finite Element software Zebulon using either an explicit or an implicit time integration scheme. Uniaxial tension tests were used to tune the model parameters for both materials, considering various values of temperature and hydrogen levels

  1. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  2. The obtainment of highly concentrated uranium pellets for plate type (MTR) fuel by dispersion of uranium aluminides in aluminium

    International Nuclear Information System (INIS)

    Morando, R.A.; Raffaeli, H.A.; Balzaretti, D.E.

    1980-01-01

    The use of the intermetallic UAl 3 for manufacturing plate type MTR fuel with 20% U 235 enriched uranium and a density of about 20 kg/m 3 is analyzed. The technique used is the dispersion of UAl 3 particles in aluminium powder. The obtainment of the UAl 3 intermetallic was performed by fusion in an induction furnace in an atmosphere of argon at a pressure of 0.7 BAR (400 mm) using an alumina melting pot. To make the aluminide powder and attain the wished granulometry a cutting and a rotating crusher were used. Aluminide powders of different granulometries and different pressures of compactation were analyzed. In each case the densities were measured. The compacts were colaminated with the 'Picture Frame' technique at temperatures of 490 and 0 deg C with excellent results from the manufacturing view point. (M.E.L.) [es

  3. Long term immersion test of aluminum alloy AA 6061 used for fuel cladding in MTR type reactors

    International Nuclear Information System (INIS)

    Linardi, Evelina M.; Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana

    2009-01-01

    In this work we present the results of long term immersion tests performed in the aluminum alloy AA 6061, used for fuel cladding in MTR type reactors. The tests were performed at open circuit potential in high purity water (ρ = 18.2 MΩ.cm) and in 10 -3 M NaCl solution. Two kinds of assemblies were studied: simple sheets and artificial crevices, immersed during 6, 12 and 18 months at room temperature. In both media and both assemblies, the aluminum hydroxide phases crystalline bayerite and bohemite were identified. It was found that a kind of localized attack named alkaline attack occurs around the iron-rich intermetallics. These particles were confirmed to control the corrosion of the AA 6061 alloy in an aerated medium. Immersion times for up to 18 months did not increase the oxide growth or the alkaline attack on the AA 6061 alloy. (author)

  4. On the calculation of flow and heat transfer characteristics for CANDU-type 19-rod fuel bundles

    International Nuclear Information System (INIS)

    Yuh-Shan Yueh; Ching-Chang Chieng

    1987-01-01

    A numerical study is reported of flow and heat transfer in a CANDU-type 19 rod fuel bundle. The flow domain of interest includes combinations of trangular, square, and peripheral subchannels. The basic equations of momentum and energy are solved with the standard k--ε model of turbulence. Isotropic turbulent viscosity is assumed and no secondary flow is considered for this steady-state, fully developed flow. Detailed velocity and temperature distributions with wall shear stress and Nusselt number distributions are obtained for turbulent flow of Re = 4.35 x 10 4 , 10 5 , 2 x 10 5 , and for laminar flow of Re--2400. Friction factor and heat transfer ceofficients of various subchannels inside the full bundle are compared with those of infinite rod arrays of triangular or square arrangements. The calculated velocity contours of peripheral subchannel agreed reasonably with measured data

  5. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 1. Summary: alternatives for the back of the LWR fuel cycle types and properties of LWR fuel cycle wastes projections of waste quantities; selected glossary

    International Nuclear Information System (INIS)

    1976-05-01

    Volume I of the five-volume report contains executive and technical summaries of the entire report, background information of the LWR fuel cycle alternatives, descriptions of waste types, and projections of waste quantities. Overview characterizations of alternative LWR fuel cycle modes are also included

  6. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

  7. Evaluation of Corrosion of the Dummy ''EE'' Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    International Nuclear Information System (INIS)

    Brower, Jeffrey Owen; Glazoff, Michael Vasily; Eiden, Thomas John; Rezvoi, Aleksey Victor

    2016-01-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and ''horseshoeing'' defects were readily observable on the surface of the several YA-type fuel elements (these are ''dummy'' plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth ''S'' curve, was represented by a series temperature rise ''humps,'' which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed sscalloping and possibly pitting degradation on the YA-M fuel elements. In the case

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Ogiya, Shunsuke.

    1989-01-01

    For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

  9. Fuel Services

    International Nuclear Information System (INIS)

    Silberstein, A.

    1982-09-01

    FRAGEMA has developed most types of inspection equipments to work on irradiated fuel assemblies and on single fuel rods during reactor outages with an efficiency compatible with the utilities operating priorities. In order to illustrate this statement, two specific examples of inspection equipments are shortly described: the on-site removable fuel rod assembly examination stand, and the fuel assembly multiple examination device. FRAGEMA has developed techniques for the identifiction of the leaking fuel rods in the fuel assembly and the tooling necessary to perform the replacement of the faulted element. These examples of methods, techniques and equipments described and the experience accumulated through their use allow FRAGEMA to qualify for offering the supply of the corresponding software, hardware or both whenever an accurate understanding of the fuel behaviour is necessary and whenever direct intervention on the assembly and associated components is necessary due to safety, operating or economical reasons

  10. Fuel spacer

    International Nuclear Information System (INIS)

    Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

    1992-01-01

    The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

  11. Experimental investigation concerning the influence of fuel type and properties on the injection and atomization of liquid biofuels in an optical combustion chamber

    International Nuclear Information System (INIS)

    Galle, J.; Defruyt, S.; Van de Maele, C.; Rodriguez, R. Piloto; Denon, Q.; Verliefde, A.; Verhelst, S.

    2013-01-01

    Due to the scarcity of fossil fuels and the future stringent emission limits, there is an increasing interest for the use of renewable biofuels in compression ignition engines. However, these fuels have different physical, chemical and thermodynamic properties affecting atomization, spray development and combustion processes. The results reported in this paper have been obtained by experimentation with a constant volume combustion chamber. The influences of physical fuel properties on injections under non-evaporating conditions are studied, using a pump-line-nozzle system from a medium speed diesel engine with injection pressures up to 1200 bar, by changing the fuel type and temperature. Experiments were conducted for diesel, biodiesel, straight vegetable oils and animal fats. Injection pressure and needle lift measurements were analyzed. A high speed camera was used to visualize the spray, which enabled us to study the spray penetration and spray angle. Our results show that the fuel temperature is an important parameter to control because it significantly affects the fuel properties. Both the injection timing and injection duration are affected by the fuel properties. The influences of these properties on the spray development were less pronounced. At low temperatures, a strongly deteriorated atomization of oils and fats was observed. -- Highlights: • Spray measurements in an optical combustion chamber. • Influence on the injections system is compared for different bio-fuels. •Temperature effects the fuel properties, with strong influence on the injection system. • Viscosity has significant influence on atomization, especially for viscous fuels. • No difference for spray penetration and angle unlike the mass distribution

  12. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  13. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

  14. Development of a toroidal shell-type shock absorber for an irradiated fuel shipping cask

    International Nuclear Information System (INIS)

    Sugita, Y.; Mochizuki, S.

    1983-01-01

    This study described the design method of a toroidal shell-type shock absorber and the dynamic responses of the cask body, the internal structure and water when this shock absorber was used. Conclusions are: the calculated results on the basis of the master curves of non-dimensionalized force-deflection relations by static compression tests show a close agreement with the experimental results; the internal structure moves together with the cask body in every position; and the maximum water pressure is larger by a factor of 1.2 than the static pressure multiplied by the maximum deceleration in every direction due to the low-frequency wave propagation

  15. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  16. Effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu K; Malachenko, L L; Medvedev, A V; Solyany, V I; Sukhanov, G I; Tonkov, V Yu

    1987-05-01

    Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. 4 references, 7 figures, 3 tables.

  17. The effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Malachenko, L.L.; Medvedev, A.V.; Solyany, V.I.; Sukhanov, G.I.; Tonkov, V.Yu.

    1987-01-01

    Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. (author)

  18. The analysis of effect of gaseous fission products on service-ability of various types of fuel elements used in research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.A.; Kulakov, G.V.; Kosaurov, A.A.; Dobrikova, I.V.; Morozov, A.A.

    2005-01-01

    The analysis of the work to license U-Mo dispersion fuel that has been carried out by program RERTR shows that at high burnups the U-Mo particles interacts with an Al matrix to form porosity in the central areas of the fuel meat. This results leads to change the loading of cladding due to the formation of long-length defects in the fuel meat. As a result, the pressure of gases leads to a larger shape change (pillowing) of fuels and in some instances to their failure. The paper presents the results of modeling pillowing process proceeding in Russia's research reactor fuels of different types (tubular and rod fuels) that is effected by the accumulation and pressure of fission gas in meat porosity. The problem has been resolved in the viscous-plasticity statement using the program MARC for finite element analysis. The behavior of fission gas (FG) in open porosity has been analyzed to reveal the influence of the fuel design on their resistance to the pressure of FG. (author)

  19. Implementation of a quality assurance system for the design and manufacturing of fuel assembly MTR-plate type

    International Nuclear Information System (INIS)

    Koll, J.H.

    1987-01-01

    Since more than 30 years ago, fuel assemblies (FA) of the MTR-Plate type, for research reactors, have been developed and produced using well known technologies, with different methods for the design, manufacturing, quality control and subsequent verification of FA behaviour, as well as of the design data. The FA and its reliability has been improved through the recycling of the obtained information. No nuclear accidents or major incidents have taken place that can be blamed to FA due to design, manufacturing or its use. Since the 70's, the use of Quality Assurance methodology has been increased, especially for Nuclear Power Plants, in order to ensure safety for these reactors. The use of QA for reactors for research, testing or other uses, has also been steadily increased, not only due to safety reasons, but also because of its convenience for a good operation, being presently a common requirement of the operator of the installation. Herewith is described the way the QA system that has been developed for the design, manufacturing, quality control and supply of MTR-plate type FA, at the Development Section of the Argentine Atomic Energy Commission (CNEA). (Author)

  20. Fuels processing for transportation fuel cell systems

    Science.gov (United States)

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  1. Power distribution investigation in the transition phase of the low moderation type MOX fueled LWR from the high conversion core to the breeding core

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

    2011-01-01

    The key concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a core transition from a high conversion (HC) type to a plutonium breeding (BR) type in a same reactor system only by replacing fuel assemblies. Consequently in this transition phase, there are two types of assemblies in the same core. Due to the differences of the two assembly types, region-wise soft to hard neutron spectra appears and result in a large power peaking. Therefore, power distribution of FLWR in the HC to BR transition phase was studied by performing assembly and core calculations. For the whole core calculation, a new 14-group energy structure is developed to better represent the power distribution obtained with the fine 107-group structure than the 9-group structure in the previous evaluations. Calculations on few assemblies geometries show large local power peakings can be effectively reduced by considering plutonium enrichment distribution in an assembly. In the whole core calculation, there is a power level mismatch between HC and BR assemblies, but overall power distribution flattening is possible by optimizing fuel assemblies loading. Although the fuel loading should be decided also taking into account the void coefficient, transition from HC to BR type FLWR seems feasible without difficulty. (author)

  2. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  3. Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

    Energy Technology Data Exchange (ETDEWEB)

    Rossa, Riccardo, E-mail: rrossa@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Borella, Alessandro, E-mail: aborella@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Labeau, Pierre-Etienne, E-mail: pelabeau@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Pauly, Nicolas, E-mail: nipauly@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Meer, Klaas van der, E-mail: kvdmeer@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium)

    2015-08-11

    The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of {sup 239}Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a {sup 239}Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to {sup 239}Pu, in comparison with a {sup 235}U fission chamber, with a {sup 3}He proportional counter, and with a {sup 10}B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the {sup 239}Pu and {sup 235}U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the {sup 3}He and {sup 10}B proportional counters to increase the sensitivity to {sup 239}Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies.

  4. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

  5. Experience with nuclear fuel utilization in Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    Harizanov, Y [Committee on the Use of Atomic Energy for Peaceful Purposes, Sofia (Bulgaria)

    1997-12-01

    The presentation on experience with nuclear fuel utilization in Bulgaria briefly reviews the situation with nuclear energy in Bulgaria and then discusses nuclear fuel performance (amount of fuel loaded, type of fuel, burnup, fuel failures, assemblies deformation). 2 tabs.

  6. Study on usage of low enriched uranium Russian type fuel elements for design of an experimental ADS research reactor

    International Nuclear Information System (INIS)

    Pesic, M.P.

    2005-01-01

    Conceptual design of an accelerator driven sub-critical experimental research reactor (ADSRR) was initiated in 1999 at the Vinca Institute of Nuclear Sciences, Serbia and Montenegro. Initial results of neutronic analyses of the proposed ADSRR-H were carried out by Monte Carlo based codes and available high-enriched uranium dioxide (HEU) dispersed Russian type TVR-S fuel elements (FE) placed in a lead matrix. Beam of charged particles (proton or deuteron) would be extracted from the high-energy channel H5B of the VINCY cyclotron of the TESLA Accelerator Installation. In 2002, the Vinca Institute has, in compliance with the Reduced Enrichment for Research and Test Reactors (RERTR) Program, returned fresh HEU TVR-S type FEs back to the Russian Federation. Since usage of HEU FEs in research reactors is not further recommended, a new study of an ADSRR-L conceptual design has initiated in Vinca Institute in last two years, based on assumed availability of low-enriched uranium (LEU) dispersed type TVR-S FEs. Initial results of numerical simulations of this new ADSRR-L, published for the first time in this paper, shows that such a small low neutron flux system can be used as an experimental - 'demonstration' - ADS with neutron characteristics similar to proposed well-known lead moderated and cooled power sub-critical ADS with intermediate neutron spectrum. Neutron spectrum characteristics of the ADSRR-L are compared to ones of the ADSRR-H with the same mass (7.7 g) of 235 U nuclide per TVR-S FE. (author)

  7. Safety aspects of long-term dry interim storage of Type B spent fuel and high-level transport casks

    International Nuclear Information System (INIS)

    Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R.

    2004-01-01

    Based on the German decision to minimise transport of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities were licensed until the end of 2003. Because of the large amount of Type B(U) transport casks which are going to be used for long-term interim storage the question of time-limited Type B(U) licence maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of qualification of transport casks for interim storage is the long-term behaviour of the metallic seal-lid system. Here we present results from current long-term experimental tests with metallic 'Helicoflex' seals in which pool water is enclosed. This series of tests has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an exchange of experience, know-how and state-of-the-art between authorities and technical experts with regard to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called 'coordinating institution for cask dispatching information' ('KOBAF') which entails management of an online database of cask-specific documents and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the coming decades. (author)

  8. Safety aspects of long-term dry interim storage of type-B spent fuel and HLW transport casks

    International Nuclear Information System (INIS)

    Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R.

    2004-01-01

    Based on the German decision to minimise transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities have been licensed till the end of 2003. Because of the large amount of type-B transport casks which are going to be used for long-term interim storage the question of time limited type-B license maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of transport cask qualification for interim storage is the long-term behaviour of the metallic seal lid system. Concerning this results from current experimental long-term tests with metallic ''Helicoflex''-seals in which pool water is enclosed are presented. The test series has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an authorities' and technical experts' exchange of experience, know-how and state of the art referring to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called ''co-ordinating institution for cask dispatching information'' (''KOBAF'') which contains an online data base and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the next decades

  9. Effect of inlet fuel type on the degradation of Ni/YSZ anode of solid oxide fuel cell by carbon deposition

    Directory of Open Access Journals (Sweden)

    Suttichai Assabumrungrat

    2006-11-01

    Full Text Available According to the high operating temperature of Solid Oxide Fuel Cell (SOFC (700-1100ºC, it is known that some hydrocarbon fuels can be directly used as inlet fuel instead of hydrogen by feeding straight to the anode. This operation is called a direct internal reforming SOFC (DIR-SOFC. However, the major difficulty of this operation is the possible degradation of anode by the carbon deposition, as the carbon species are easily formed. In the present work, the effect of inlet fuel (i.e. H2, synthesis gas (H2+CO, CH4, CH4+H2O, CH3OH+H2O, and C2H5OH+H2O on the degradation of nickel cermet (Ni/YSZ, which is the most common anode material of SOFC, was studied.It was found from the work that hydrogen and synthesis gas (CO+H2 are proper to be used as direct inlet fuels for DIR-SOFC with Ni/YSZ anode, since the carbon formation on Ni/YSZ occurred in the small quantity. The mixture of methane and steam (CH4+H2O can also be used as the inlet feed, but the H2O/CH4 ratio plays an important role. In contrast, pure methane (CH4, methanol with steam (CH3OH+H2O and ethanol with steam (C2H5OH+H2O are not suitable for using as direct inlet fuel for DIR-SOFC with Ni/YSZ anode even the higher H2O/CH3OH and H2O/C2H5OH ratios were applied.

  10. Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel; Metodo Optimizado de Extraccion por Ultrasonidos para la Determinacion de PAHs Seleccionados en un Suelo Contaminado con Fuel de Tipo II

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

    2010-10-21

    A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Bando, Masaru.

    1993-01-01

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Yokota, Tokunobu.

    1990-01-01

    A fuel assembly used in a FBR type nuclear reactor comprises a plurality of fuel rods and a moderator guide member (water rod). A moderator exit opening/closing mechanism is formed at the upper portion of the moderator guide member for opening and closing a moderator exit. In the initial fuel charging operation cycle to the reactor, the moderator exit is closed by the moderator exit opening/closing mechanism. Then, voids are accumulated at the inner upper portion of the moderator guide member to harden spectrum and a great amount of plutonium is generated and accumulated in the fuel assembly. Further, in the fuel re-charging operation cycle, the moderator guide member is used having the moderator exit opened. In this case, voids are discharged from the moderator guide member to decrease the ratio, and the plutonium accumulated in the initial charging operation cycle is burnt. In this way, the fuel economy can be improved. (I.N.)

  13. Bio-based production of fuels and industrial chemicals by repurposing antibiotic-producing type I modular polyketide synthases: opportunities and challenges

    DEFF Research Database (Denmark)

    Yuzawa, Satoshi; Keasling, Jay D.; Katz, Leonard

    2017-01-01

    Complex polyketides comprise a large number of natural products that have broad application in medicine and agriculture. They are produced in bacteria and fungi from large enzyme complexes named type I modular polyketide synthases (PKSs) that are composed of multifunctional polypeptides containin...... have applications as fuels or industrial chemicals....

  14. Effects of vehicle ventilation system, fuel type, and in-cabin smoking on the concentration of toluene and ethylbenzene in Pride cars

    Directory of Open Access Journals (Sweden)

    Masoud Rismanchian

    2013-01-01

    Conclusion: The ventilation condition, fuel type, and in-cabin smoking were not significantly impressive on the toluene and ethylbenzene concentrations inside the cars. However, simultaneous usage of the vehicle ventilation system and natural ventilation (windows could lead to little decrease in toluene concentration levels inside the car, while smoking consumption by passengers can increase them.

  15. Reprocessing facility for spent fuel from LWR type reactors and mixed-oxide fuel fabrication plant in the Taxoeldern Forest near Wackersdorf, Bavaria (WAA) - first partial licence

    International Nuclear Information System (INIS)

    1985-01-01

    Full text of the first partial licence for the WAA, allowing erection of the following buildings or structures: External fence; guardhouse 1, i.e. the building and the ground connection system with lightning protection system, the fire alarm system and mobile fire-fighting systems; the fuel receiving station, including building and operation systems; excavation work for the main reprocessing building. (HP) [de

  16. A Favela Santa Marta e seus guias de turismo: identidade, mobilização e conflito

    Directory of Open Access Journals (Sweden)

    Gabriel Ferreira Barbosa

    2015-03-01

    Full Text Available A proposta deste artigo é descrever o processo de constituição de um mercado turístico na Favela Santa Marta e os efeitos observados no que se refere aos conflitos e consequentes mobilizações que reivindicam a legitimidade de se contar a história desse lugar. Por ter sido a primeira favela da cidade do Rio de Janeiro onde foi instalada uma Unidade de Polícia Pacificadora (UPP em dezembro de 2008[1], a Santa Marta passa a ser reconhecida como uma “favela modelo” por parte da administração pública, como se ela fosse um “laboratório” para experiências de políticas públicas. Esse processo ocorre em um contexto – notadamente os megaeventos esportivos como a Copa do Mundo FIFA 2014 e as Olimpíadas de 2016 -  em que se busca um “reencantamento” das representações sobre as favelas, mais especificamente aquelas onde as UPPs passam a fazer parte do cotidiano de seus moradores. O programa Rio Top Tour, que é a política analisada aqui, foi lançado na Favela Santa Marta, e consistiu na formação de guias turísticos locais e estímulo ao desenvolvimento de um mercado turístico. A partir da atuação desses guias e a sua mobilização para serem os legítimos portadores da narrativa sobre a história da Favela Santa Marta no âmbito do mercado turístico local, analisarei os conflitos que resultam desse processo. Antes disso, entretanto, é preciso apresentar este lugar e as mudanças que vêm ocorrendo nos últimos anos desde a instalação da UPP – que é o corte temporal proposto neste trabalho.

  17. Accident analysis for new reactor concepts and VVER type reactor design with advanced fuel. STC with Russia. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.

    2000-10-01

    In the frame of a project on scientific-technical cooperation funded by BMBF/BMWi, the 3D reactor dynamics code DYN3D developed at Forschungszentrum Rossendorf (FZR), has been transferred to the Institute of Physics and Power Engineering (IPPE) Obninsk in Russia and integrated into the software package of IPPE. DYN3D has been coupled to a thermohydraulic system code used in IPPE making available 3D neutron kinetics within this software package. A new macroscopic cross section library has been created using a modified version of the WIMS/D4 code. This library includes data for modernized fuel design containing burnable absorbers in different concentrations, which is tested in VVER-1000 type reactors. The cross section library has been connected to DYN3D. Calculations were performed to check the library in comparison with other data libraries and codes. The code DYN3D and the coupled 3D neutron kinetics/thermal hydraulics code system were used to perform analyses of Anticipated Transients Without Scram (ATWS) for the reactor design ABV-67, an integral reactor concept with small power developed under participation of IPPE. The fluid dynamics code DINCOR developed at IPPE was transferred to FZR. It was used in validation calculations on test problems for the short-term core melt behaviour (CORVIS experiments). (orig.) [de

  18. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1985-02-01

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation [fr

  19. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    International Nuclear Information System (INIS)

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  20. Modeling CANDU type fuel behaviour during extended burnup irradiations using a revised version of the ELESIM code

    International Nuclear Information System (INIS)

    Arimescu, V.I.; Richmond, W.R.

    1992-05-01

    The high-burnup database for CANDU fuel, with a variety of cases, offers a good opportunity to check models of fuel behaviour, and to identify areas for improvement. Good agreement of calculated values of fission-gas release, and sheath hoop strain, with experimental data indicates that the global behaviour of the fuel element is adequately simulated by a computer code. Using, the ELESIM computer code, the fission-gas release, swelling, and fuel pellet expansion models were analysed, and changes made for gaseous swelling, and diffusional release of fission-gas atoms to the grain boundaries. Using this revised version of ELESIM, satisfactory agreement between measured values of fission-gas release was found for most of the high-burnup database cases. It is concluded that the revised version of the ELESIM code is able to simulate with reasonable accuracy high-burnup as well as low-burnup CANDU fuel

  1. Neutronic analysis of HEU to LEU conversion calculation for AEOI 5 MW pool-type MTR fuel research reactor core

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Lutz, D.; Bartsch, G.

    1987-07-01

    The possibility of converting HEU(93%) fuel to LEU(20%) fuel without or with slight alteration to the fuel element geometry is discussed. The fuel density varies between 1.7 to 4.1 g U-235/cm. In cross section generation a unit cell with an extra zone to account for extra Al and water was considered. In burnup calculations a sequential shuffling pattern was assumed with fixed position control fuel elements. A cross section data set in 45 energy groups were generated using RSYST/CGM system using the cross section library JFET. Then for 2D-diffusion calculations homogenized and condensed 5 energy group cross sections were prepared. (orig./HP)

  2. Sistemática para avaliação da conformidade: uma proposta para uso de cão-guia

    OpenAIRE

    Brito, Joana D'Arc de

    2009-01-01

    O objetivo do presente estudo é identificar e propor quais os principais requisitos para a Avaliação dos Centros de Treinamento, de Treinadores e dos Instrutores Autônomos de cães-guia, de forma a atender o que prescreve a Lei n.º 11.126, de 27 de junho de 2005 no concernente às necessidades de adaptação dos segmentos envolvidos. Esta pesquisa foi estruturada em 3 (três) etapas. No primeiro momento utilizou-se de fontes bibliográficas para o levantamento de obras de referência para o desenvol...

  3. Benchmark for the qualification of gamma shielding calculation methods for light-water type reactor spent fuels

    International Nuclear Information System (INIS)

    Blum, P.; Cagnon, R.; Nimal, J.C.

    1982-01-01

    This report gives the results of a campaign of gamma dose rates measurement in the vicinity of a transport package loaded with 12 PWR spent fuel assemblies, so that the characteristics of the package and the fuel. It describes the measuring methods, and gives the accuracy of the data which will be usefull, as benchmarks, to the control of the calculation methods used to verify the gamma shielding of the packages. It shows how to calculate gamma dose rates from the data given on the package and the fuel, and gives the results of a calculation with the Mecure IV code and compares them to the measurements

  4. Solid oxide fuel cells with apatite-type lanthanum silicate-based electrolyte films deposited by radio frequency magnetron sputtering

    Science.gov (United States)

    Liu, Yi-Xin; Wang, Sea-Fue; Hsu, Yung-Fu; Wang, Chi-Hua

    2018-03-01

    In this study, solid oxide fuel cells (SOFCs) containing high-quality apatite-type magnesium doped lanthanum silicate-based electrolyte films (LSMO) deposited by RF magnetron sputtering are successfully fabricated. The LSMO film deposited at an Ar:O2 ratio of 6:4 on an anode supported NiO/Sm0.2Ce0·8O2-δ (SDC) substrate followed by post-annealing at 1000 °C reveals a uniform and dense c-axis oriented polycrystalline structure, which is well adhered to the anode substrate. A composite SDC/La0·6Sr0·4Co0·2Fe0·8O3-δ cathode layer is subsequently screen-printed on the LSMO deposited anode substrate and fired. The SOFC fabricated with the LSMO film exhibits good mechanical integrity. The single cell with the LSMO layer of ≈2.8 μm thickness reports a total cell resistance of 1.156 and 0.163 Ωcm2, open circuit voltage of 1.051 and 0.982 V, and maximum power densities of 0.212 and 1.490 Wcm-2 at measurement temperatures of 700 and 850 °C, respectively, which are comparable or superior to those of previously reported SOFCs with yttria stabilized zirconia electrolyte films. The results of the present study demonstrate the feasibility of deposition of high-quality LSMO films by RF magnetron sputtering on NiO-SDC anode substrates for the fabrication of SOFCs with good cell performance.

  5. Study on high conversion type core of innovative water reactor for flexible fuel cycle (FLWR) for minor actinide (MA) recycling

    International Nuclear Information System (INIS)

    Fukaya, Yuji; Nakano, Yoshihiro; Okubo, Tsutomu

    2009-01-01

    In order to ensure sustainable energy supplies in the future based on the well-established light water reactor (LWR) technologies, conceptual design studies have been performed on the innovative water reactor for flexible fuel cycle (FLWR) with the high conversion ratio core. For early introduction of FLWR without a serious technical gap from the LWR technologies, the conceptual design of the high conversion type one (HC-FLWR) was constructed to recycle reprocessed plutonium. Furthermore, an investigation of minor actinide (MA) recycling based on the HC-FLWR core concept has been performed and is presented in this paper. Because HC-FLWR is a near-term technology, it would be a good option in the future if HC-FLWR can recycle MAs. In order to recycle MAs in HC-FLWR, it has been found that the core design should be changed, because the loaded MA makes the void reactivity coefficient worse and decreases the discharge burn-up. To find a promising core design specification, the investigation on the core characteristics were performed using the results from parameter surveys with core burn-up calculations. The final core designs were established by coupled three dimensional neutronics and thermal-hydraulics core calculations. The major core specifications are as follows. The plutonium fissile (Puf) content is 13 wt%. The discharge burn-up is about 55 GWd/t. Around 2 wt% of Np or Am can be recycled. The MA conversion ratios are around unity. In particular, it has been found that loaded Np can be transmuted effectively in this core concept. Therefore, these concepts would be a good option to reduce environmental burdens.

  6. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  7. Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

    International Nuclear Information System (INIS)

    Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

    2003-07-01

    Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

  8. Properties, degradation and high temperature fuel cell test of different types of PBI and PBI blend membranes

    DEFF Research Database (Denmark)

    Li, Qingfeng; Rudbeck, Hans Christian; Chromik, Andreas

    2010-01-01

    Polybenzimidazoles (PBIs) with synthetically modified structures and their blends with a partially fluorinated sulfonated aromatic polyether have been prepared and characterized for high temperature proton exchange membrane fuel cells. Significant improvement in the polymer chemical stability...

  9. Fuel consumption: short term and long term price impacts per population type; Consommation de carburant: effets des prix a court et a long termes par type de population

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    This report presents assessments of the price sensitivity of household fuel consumption. After a literature review on price-elasticity assessments and the use of pseudo-panels, the investigation analyses the deciding factors of the household fuel expense and its evolution between 1985 and 2006. It proposes a short term price-elasticity assessment based on the most recent survey, and also proposes price-elasticity assessments for sub-populations, notably in terms of income level or location (rural or urban areas)

  10. Recent status and future aspect of plate type fuel element technology with high uranium density at NUKEM

    International Nuclear Information System (INIS)

    Hrovat, M.F.; Hassel, H.-W.

    1983-01-01

    According to the present state of development full size test fuel elements with UAl x , U 3 O 8 , and U 3 Si 2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cc for UAI x , 2.9 g/cc for U 3 O 8 , and 4.76 g/cc for U 3 Si 2 . The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UA1 x , 60% with U 3 O 8 , 39% with U 3 Si 2 ; no defects could be detected. For an economical fuel element production with reduced 235-U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAl x and U 3 O 8 fuel were successfully irradiated in the ORR up to a burnup of 75 %. The relatively high uranium meat densities of these test plates amount to 2.2 g/cc for UAI x , and 3.14 g/cc for U 3 O 8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted. (author)

  11. A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

    International Nuclear Information System (INIS)

    Yoshisaki, Magno B.; Leopando, Leonardo S.

    1999-03-01

    Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

  12. Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties. (author)

  13. Status and development potential of proven reactor types and fuel cycles, and their role in a medium-to-long range energy supply strategy

    International Nuclear Information System (INIS)

    Maerkl, H.

    1982-01-01

    After a general review of the present world-wide energy situation (with particular reference to those of the Federal Republic of Germany and of Argentina) the possible contribution of nuclear energy in general, and of proven light water and heavy water reactor types in particular, to meeting the energy demand is discussed. The technical and economic development potential of those reactors is evaluated, both regarding plant components technology as well as fuel and fuel cycle improvement, with special emphasis on the Pressure Vessel Heavy Water Reactor type. The last section presents some results of nuclear reactor strategy calculations made for a scenario similar to that of Argentina over the period from 1970 through 2040 and involving the use of: A) heavy water reactors (HWR's) only, with and without plutonium recycling, and B) the use of HWR's plus fast breeder reactors. (M.E.L.) [es

  14. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  16. Development and testing of a hybrid system with a sub-kW open-cathode type PEM (proton exchange membrane) fuel cell stack

    International Nuclear Information System (INIS)

    Huang, Zhen-Ming; Su, Ay; Liu, Ying-Chieh

    2014-01-01

    In this study, the performance of a polymer electrolyte membrane fuel cell stack has been evaluated for a hybrid power system test platform. To simulate vehicle acceleration, the stack was operated under dynamic-loading, and to demonstrate the exchange of power flow between two power sources the hybrid power system was tested under three different modes. A unit cell was fabricated for high stack performance and the stack was constructed with 18 open-cathode type fuel cells. Air which acts as a coolant as well as an oxidant for electrochemical reactions is provided by a pair of fans. The capabilities of the stack for hybrid power system test platform were validated by successful dynamic-loading tests. The performance of the stack for various air fan voltage was evaluated and an optimal value was concluded. The conditions like inlet temperature of H 2 and the stack current were established for maximum power. It was also found that humidification of hydrogen at anode inlet degrades the stack performance and stability due to flooding. Evidence shows that for the higher overall performance, the fuel cell acts continuously on constant current output. The study contributes to the design of mobility hybrid system to get better performance and reliability. - Highlights: • An open-cathode type PEMFC (polymer electrolyte membrane fuel cell) stack (rated output 300 W) was fabricated. • The open-cathode configuration simplifies the design of a stack system. • Assess the feasibility of combining a fuel cell stack in a hybrid system. • The study contributes to the design of mobility hybrid system to get better performance and reliability

  17. Constant strength fuel-fuel cell

    International Nuclear Information System (INIS)

    Vaseen, V.A.

    1980-01-01

    A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

  18. The Influence of Particle Size, Fluidization Velocity, and Fuel Type on Ash-Induced Agglomeration in Biomass Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Gatternig, Bernhard, E-mail: bernhard.gatternig@cbi.uni-erlangen.de; Karl, Jürgen [Chair of Energy Process Engineering, Friedrich-Alexander University Erlangen-Nürnberg, Nuremberg (Germany)

    2014-11-19

    Agglomeration of the bed material is one of the main obstacles for biomass utilization in fluidized bed combustors. Especially, high-potential fuels such as fast growing energy crops or biogeneous residues are affected because of their high content of alkaline metals. Despite ongoing research efforts, the knowledge base on what fuels are affected is still limited. This paper describes the design and installation of two lab-scale reactors for the experimental determination of agglomeration temperatures. The reactor concept and measurement method were developed under consideration of experiences from existing test rigs published in literature. Preliminary tests confirmed a reproducibility of ±5°C for both new reactors. The results of an extended measurement campaign (156 test runs of 25 fuel species at a wide range of the operational parameters “bed particle size,” “gas velocity,” and “bed ash accumulation”), based on “design of experiment” (DoE) criteria, showed high-agglomeration tendencies for residues (e.g., dried distillery grains, corn cobs) while woody energy crops (e.g., willow, alder) exhibited very stable combustion behavior. The operating parameters influenced the agglomeration behavior to a lesser degree than different ash compositions of fuel species tested. An interpolation within the DoE factor space allowed for a subsequent comparison of our results with experiments reported in literature. Good agreement was reached for fuels of comparable ash composition considering the interpolation errors of ±32°C on average.

  19. O simbolismo da águia na religiosidade nórdica pré-cristã e cristã

    Directory of Open Access Journals (Sweden)

    Johnni Langer

    2015-10-01

    Full Text Available Presente desde a Pré-História até o fim da Idade Média, as representações de águias no mundo nórdico assumiram diversos simbolismos religiosos. Em busca de compreender as diferentes funções deste animal em diferentes períodos, abordaremos o simbolismo da águia de forma distinta nos períodos Pré-Viking, na Era Viking e no Cristianismo da Escandinávia Medieval. Para tanto, realizaremos uma análise de diversas fontes iconográficas, arqueológicas e literárias, como jóias, monumentos e manuscritos, em um diálogo com autores clássicos — Hilda Davidson, James Graham-Campbell e Régis Boyer, por exemplo —, e com estudos atualizados de especialistas no assunto, tais como Anne-Sophie Gräslund, Jens Peter Schjødt e Kristina Jennbert. Como referenciais teóricos e metodológicos, adotamos o conceito de longa duração na abordagem de Lotte Hedeager e o conceito de símbolo para a Arqueologia das Religiões, aplicados à religiosidade nórdica.

  20. Small Engine Repair Modules (Workbook) = Reparacion de Motores Pequenos (Guia de Trabajo)

    Science.gov (United States)

    New York State Dept. of Correctional Services, Albany.

    This package contains an English-Language set of task procedure sheets dealing with small-engine repair and a Spanish translation of the same material. Addressed in the individual sections of the manual are the following aspects of engine tune-up, reconditioning, and troubleshooting: servicing air cleaners; cleaning gas tanks, fuel lines, and fuel…

  1. Remote fabrication of (Th, {sup 233}U)O{sub 2} pellet-type fuels for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A

    1981-05-15

    Thorium fuels enriched with {sup 233}U must be fabricated in shielded cells because of high gamma and alpha activity. A conceptual design of a remotely operated plant to produce gamma-active pellet fuels has been made. The plant consists of eight fabrication canyons, two repair canyons, and several miscellaneous cells. Process equipment is modular, easily disconnected, and mounted on plates for easy removal. Equipment consists of a combination of robotics, hard automation, and conventional process equipment. The plant is operated from a central control room with the assistance of a sophisticated computer-based control and information system. Many of the automated process steps are preprogrammed on the control computer and executed on demand by the supervising operator. The technology to build such a plant exists today but needs to be adapted to the needs of the recycle fuel industry. (author)

  2. The analysis of coolant-velocity distribution in plat-typed fuel element using CFD method for RSG-GAS research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Darwis Isnaini; Endiah Puji Hastuti

    2013-01-01

    The measurement experiment for coolant-velocity distribution in the subchannel of fuel element of RSG-GAS research reactor is difficult to be carried out due to too narrow channel and subchannel placed inside the fuel element. Hence, the calculation is required to predict the coolant-velocity distribution inside subchannel to confirm that the handle presence does not ruin the velocity distribution into every subchannel. This calculation utilizes CFD method, which respect to 3-dimension interior. Moreover, the calculation of coolant-velocity distribution inside subchannel was not ever carried out. The research object is to investigate the distribution of coolant-velocity in plat-typed fuel element using 3-dimension CFD method for RSG-GAS research reactor. This research is required as a part of the development of thermalhydraulic design of fuel element for innovative research reactor as well. The modeling uses ½ model in Gambit software and calculation uses turbulence equation in FLUENT 6.3 software. Calculation result of 3D coolant-velocity in subchannel using CFD method is lower about 4.06 % than 1D calculation result due to 1D calculation obeys handle availability. (author)

  3. Polyester synthesis for application in PEMFC type fuel cells; Sintese de poliester para aplicacao em celulas a combustivel do tipo PEM

    Energy Technology Data Exchange (ETDEWEB)

    Fiuza, R.P.; Souza, D.R. de; Fiuza, R.A.; Jose, N.M.; Boaventura, J.S. [Universidade Federal da Bahia (UFBA), Salvador, BA (Brazil). Inst. de Quimica], e-mail: raigenis@gmail.com

    2006-07-01

    The PEMFC (Proton Exchange Membrane Fuel Cell), along the SOFC (Solid Oxide Fuel Cell), is the most important technology, among the various types of fuels cell. The PEMFC shows a large versatility of applications, both for stationary and mobile use. However the PEMFC presents high manufacture cost, directly impacting in the cost of the produced energy. This work contemplates the previews sulfonation of phtalic acid and its subsequent polymerization with glycerol, using as catalytic tin dibutyl-dilaurate. The obtained material has been characterized by DSC, TGA, FTIR, MEV, DRX and XRF. The gotten results indicated that phtalic acid was sulfonated and the increase of the sulfonation degree significantly increased the crystallinity of the sulfonated ftalico acid. Furthermore, the polymer produced from the sulfonated monomer presented adequate thermal resistance and a high content of conducting groups, necessary conditions for application as electrolyte in PEMFC. All these characteristics, particularly the low cost of the reagents and the ease of production process, make the sulfonated polyester membrane a promising candidate as fuel cell electrolyte. (author)

  4. Development and implementation of computational geometric model for simulation of plate type fuel fabrication process with microspheres dispersed in metallic matrix

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Reis, Sergio C.; Braga, Daniel M.; Santos, Armindo; Ferraz, Wilmar B.

    2005-01-01

    In this report it is presented the development of a geometric model to simulate the plate type fuel fabrication process with fuels microspheres dispersed in metallic matrix, as well as its software implementation. The developed geometric model encloses the steps of pellets pressing and sintering, as well as the plate rolling passes. The model permits the simulation of structures, where the values of the various variables of the fabrication processes can be studied and modified. The following variables were analyzed: microspheres diameters, density of the powder/microspheres mixing, microspheres density, fuel volume fraction, sintering densification, and rolling passes number. In the model implementation, which was codified in DELPHI programming language, systems of structured analysis techniques were utilized. The structures simulated were visualized utilizing the AutoCAD applicative, what permitted to obtain planes sections in diverse directions. The objective of this model is to enable the analysis of the simulated structures and supply information that can help in the improvement of the dispersion microspheres fuel plates fabrication process, now in development at CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) in cooperation with the CTMSP (Centro Tecnologico da Marinha em Sao Paulo). (author)

  5. Recent status of development and irradiation performance for plate type fuel elements with reduced 235U enrichment at NUKEM

    International Nuclear Information System (INIS)

    Hrovat, M.F.; Hassel, H.W.

    1984-01-01

    According to the present state of development full size test fuel elements with the maximum uranium densities of 2,2 g U/cm 3 meat for UAlsub(x), 3,2 g U/cm 3 meat for U 3 O 8 and 4,8 g U/cm 3 meat for U 3 Si 2 can be fabricated at NUKEM in production scale. Special chemical procedures for the uranium recovery were developed ensuring an economic fuel fabrication process. The post irradiation examinations (PIE) of 12 UAlsub(x) (U density 2,2 g U/cm 3 meat) and U 3 O 8 (up to 3,1 g U/cm 3 meat) test plates irradiated in the ORR, Oak Ridge research reactor, were terminated. All 12 test plates show unobjectionable irradiation behavior. Extensive irradiation tests on full size fuel elements were performed. All inserted elements show perfect irradiation behavior. The PIE of the first HFR Petten U 3 O 8 fuel elements are in progress. The full size ORR U 3 Si 2 fuel elements with so far highest uranium density of 4,76 g U/cm 3 meat achieved a burnup of 50 % loss of 235 U up to May 1983. One element was withdrawn from the reactor for PIE, the second will be irradiated to a burnup of 75 % loss of 235 U. The further development is concentrated on Usub(x)Sisub(y) fuel with highest uranium density. U 3 Si miniplates with up to 6,1 g U/cm 3 meat are supplied meeting the required specification, U 3 Si miniplates with 6,7 g U/cm 3 are in fabrication. (author)

  6. Contribution to fuel depletion study in PWR type reactors, reactor core with three and four regions of enrichment

    International Nuclear Information System (INIS)

    Teixeira, M.C.C.

    1977-03-01

    The main methods for calculation of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Power Plant are presented and discussed. (author)

  7. Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

    International Nuclear Information System (INIS)

    Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

    2013-01-01

    Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

  8. Preparation of U-Si/U-Me (Me = Fe, Ni, Mn) aluminum-dispersion plate-type fuel (miniplates) for capsule irradiation

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro; Itoh, Akinori; Akabori, Mitsuo

    1993-06-01

    Details of equipment installed, method adopted and final products were described on the preparation of uranium silicides and other fuels for capsule irradiation. Main emphasis was placed on the preparation of laboratory-scale aluminum-dispersion plate-type fuel (miniplates) loaded to the first and second JMTR silicide capsules. Fuels contained in the capsules are as follows: (A) uranium-silicide base alloys U 3 Si 2 , Mo- added U 3 Si 2 , U 3 Si 2 +U 3 Si, U 3 Si 2 +USi, U 3 Si, U 3 (Si 0.8 Ge 0.2 ), U 3 (Si 0.6 Ge 0.4 ) (B) U 6 Me-type alloys with higher uranium density U 6 Mn, U 6 Ni, U 6 (Fe 0.4 Ni 0.6 ), U 6 (Fe 0.6 Mn 0.4 ) The powder-metallurgical picture-frame method was adopted and laboratory-scale technique was established for the preparation of miniplates. As a result of inspection for capsule irradiation, miniplates were prepared to meet the requirements of specification. (author)

  9. Developing Models to Predict the Number of Fire Hotspots from an Accumulated Fuel Dryness Index by Vegetation Type and Region in Mexico

    Directory of Open Access Journals (Sweden)

    D. J. Vega-Nieva

    2018-04-01

    Full Text Available Understanding the linkage between accumulated fuel dryness and temporal fire occurrence risk is key for improving decision-making in forest fire management, especially under growing conditions of vegetation stress associated with climate change. This study addresses the development of models to predict the number of 10-day observed Moderate-Resolution Imaging Spectroradiometer (MODIS active fire hotspots—expressed as a Fire Hotspot Density index (FHD—from an Accumulated Fuel Dryness Index (AcFDI, for 17 main vegetation types and regions in Mexico, for the period 2011–2015. The AcFDI was calculated by applying vegetation-specific thresholds for fire occurrence to a satellite-based fuel dryness index (FDI, which was developed after the structure of the Fire Potential Index (FPI. Linear and non-linear models were tested for the prediction of FHD from FDI and AcFDI. Non-linear quantile regression models gave the best results for predicting FHD using AcFDI, together with auto-regression from previously observed hotspot density values. The predictions of 10-day observed FHD values were reasonably good with R2 values of 0.5 to 0.7 suggesting the potential to be used as an operational tool for predicting the expected number of fire hotspots by vegetation type and region in Mexico. The presented modeling strategy could be replicated for any fire danger index in any region, based on information from MODIS or other remote sensors.

  10. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  11. Mathematical model of water transport in Bacon and alkaline matrix-type hydrogen-oxygen fuel cells

    Science.gov (United States)

    Prokopius, P. R.; Easter, R. W.

    1972-01-01

    Based on general mass continuity and diffusive transport equations, a mathematical model was developed that simulates the transport of water in Bacon and alkaline-matrix fuel cells. The derived model was validated by using it to analytically reproduce various Bacon and matrix-cell experimental water transport transients.

  12. Fuel cells

    International Nuclear Information System (INIS)

    Niederdoeckl, J.

    2001-01-01

    Europe has at present big hopes on the fuel cells technology, in comparison with other energy conversion technologies, this technology has important advantages, for example: high efficiency, very low pollution and parallel use of electric and thermal energy. Preliminary works for fuel cells developing and its commercial exploitation are at full speed; until now the European Union has invested approx. 1.7 billion Schillings, 60 relevant projects are being executed. The Austrian industry is interested in applying this technique to drives, thermal power stations and the miniature fuel cells as replacement of batteries in electronic products (Notebooks, mobile telephones, etc.). A general description of the historic development of fuel cells including the main types is given as well as what is the situation in Austria. (nevyjel)

  13. Influence of the silicon content on the core corrosion properties of dispersion type fuel plates; Influencia del Contenido en silicio sobre la corrosion acuosa de los nucleos de placas combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Calvo, C; Saenz de Tejada, L M; Diaz Diaz, J

    1969-07-01

    A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI{sub 3} and AI{sub 2}O{sub 3} according to the reaction. (Author)

  14. Kitchen PM2.5 concentrations and child acute lower respiratory infection in Bhaktapur, Nepal: The importance of fuel type.

    Science.gov (United States)

    Bates, Michael N; Pokhrel, Amod K; Chandyo, Ram K; Valentiner-Branth, Palle; Mathisen, Maria; Basnet, Sudha; Strand, Tor A; Burnett, Richard T; Smith, Kirk R

    2018-02-01

    Globally, solid fuels are used by about 3 billion people for cooking and a smaller number use kerosene. These fuels have been associated with acute lower respiratory infection (ALRI) in children. Previous work in Bhaktapur, Nepal, showed comparable relationships of biomass and kerosene cooking fuels with ALRI in young children, compared to those using electricity for cooking. We examine the relationship of kitchen PM 2.5 concentrations to ALRI in those households. ALRI cases and age-matched controls were enrolled from a cohort of children 2-35 months old. 24-h PM 2.5 was measured once in each participant's kitchen. The main analysis was carried out with conditional logistic regression, with PM 2.5 measures specified both continuously and as quartiles. In the kitchens of 393 cases and 431 controls, quartiles of increasing PM 2.5 concentration were associated with a monotonic increase in odds ratios (OR): 1.51 (95% CI: 1.00, 2.27), 2.22 (1.47, 3.34), 2.48 (1.63, 3.77), for the 3 highest exposure quartiles. The general kitchen concentration-response shape across all stoves was supralinear. There was evidence for increased risk with biomass stoves, but the slope for kerosene stoves was steeper, the highest quartile OR being 5.36 (1.35, 21.3). Evidence for increased risk was also found for gas stoves. Results support previous reports that biomass and kerosene cooking fuels are both ALRI risk factors, but suggests that PM 2.5 from kerosene is more potent on a unit mass basis. Further studies with larger sample sizes and preferably using electricity as the baseline fuel are needed. Copyright © 2017 Elsevier Inc. All rights reserved.

  15. Dry syngas purification process for coal gas produced in oxy-fuel type integrated gasification combined cycle power generation with carbon dioxide capturing feature.

    Science.gov (United States)

    Kobayashi, Makoto; Akiho, Hiroyuki

    2017-12-01

    Electricity production from coal fuel with minimizing efficiency penalty for the carbon dioxide abatement will bring us sustainable and compatible energy utilization. One of the promising options is oxy-fuel type Integrated Gasification Combined Cycle (oxy-fuel IGCC) power generation that is estimated to achieve thermal efficiency of 44% at lower heating value (LHV) base and provide compressed carbon dioxide (CO 2 ) with concentration of 93 vol%. The proper operation of the plant is established by introducing dry syngas cleaning processes to control halide and sulfur compounds satisfying tolerate contaminants level of gas turbine. To realize the dry process, the bench scale test facility was planned to demonstrate the first-ever halide and sulfur removal with fixed bed reactor using actual syngas from O 2 -CO 2 blown gasifier for the oxy-fuel IGCC power generation. Design parameter for the test facility was required for the candidate sorbents for halide removal and sulfur removal. Breakthrough test was performed on two kinds of halide sorbents at accelerated condition and on honeycomb desulfurization sorbent at varied space velocity condition. The results for the both sorbents for halide and sulfur exhibited sufficient removal within the satisfactory short depth of sorbent bed, as well as superior bed conversion of the impurity removal reaction. These performance evaluation of the candidate sorbents of halide and sulfur removal provided rational and affordable design parameters for the bench scale test facility to demonstrate the dry syngas cleaning process for oxy-fuel IGCC system as the scaled up step of process development. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  17. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  18. Indoor exposures to particulate matter emissions in various types of households using different cooking fuels in rural areas of south India

    Science.gov (United States)

    Deepthi, Y.; Nagendra, S. S.; Gummadi, S. N.

    2017-12-01

    Exposure to Particulate Matter (PM) that are typically generated from heavy biomass usage in cooking and from unpaved roads is a major health risk in the rural areas of developing countries. To understand the exposure levels in such areas, PM (PM10, PM2.5 and PM1) characterizations was carried out through indoor monitoring in a rural site of south India with varied cooking fuels such as only biomass, biomass plus LPG and only LPG in different types of housing namely indoor kitchen without partition (IKWO), indoor kitchen with partition (IKWP), separate enclosed kitchen outside house (SEKO) and open kitchen (OK). Results indicated that use of biomass resulted in the highest PM10 concentrations of 179.51±21µg/m3 followed by combination of biomass and LPG (101.99±21 µg/m3) and LPG (77.48±9µg/m3). Similar patterns were observed in PM2.5 and PM1 with highest emissions from biomass burning. The PM concentrations of biomass households and combination of biomass and LPG households were 233.7 % and 80.2 % respectively higher than those using cleaner fuels (LPG). The monitoring also revealed that kitchen configuration is an important determinant for indoor exposures especially for biomass households. Among biomass users, average PM10, PM2.5 and PM1 concentrations in all type of houses were above the human permissible limit with IKWP having highest concentrations followed by IKWO>SEKO>OK. Thus, biomass household have high concentrations compared to LPG because of nature of combustion of solid biomass. Also, PM concentrations were higher in enclosed indoor kitchens (IKWO and IKWP) compared to SEKO and OK type kitchen configurations. It is evident from above discussions that type of fuel and kitchen setups are major attributes impacting Indoor air pollution (IAP) in rural areas and any policy intervention to minimize IAP must give due consideration to these two factors.

  19. Development of destructive methods of burn-up determination and their application on WWER type nuclear fuels

    International Nuclear Information System (INIS)

    Hermann, A.; Stephan, H.; Nebel, D.

    1984-03-01

    Results are described of a cooperation between the Central Institute of Nuclear Research Rossendorf and the Radium Institute 'V.G. Chlopin' Leningrad in the field of destructive burn-up determination. Laboratory methods of burn-up determination using the classical monitors 137 Cs, 106 Ru, 148 Nd and isotopes of heavy metals (U, Pu) as well as the usefulness of 90 Sr, stable isotopes of Ru and Mo as monitors are dealt with. The analysis of the fuel components uranium (spectrophotometry, potentiometric titration, mass-spectrometric isotope dilution) and plutonium (spectrophotometry, coulometric titration, mass- and alpha-spectrometric isotope dilution) is fully described. Possibilities of increasing the reproducibility (automatic adjusting of measurement conditions) and the sensibility (ion impuls counting) of mass-spectrometric measurements are proposed and applied to a precise determination of Am and Cm isotopic composition. The methods have been used for burn-up analysis of spent WWER (especially WWER-440) fuel. (author)

  20. Sustainable Power Generation in Continuous Flow Microbial Fuel Cell Treating Actual Wastewater: Influence of Biocatalyst Type on Electricity Production

    Directory of Open Access Journals (Sweden)

    Zainab Z. Ismail

    2013-01-01

    Full Text Available Microbial fuel cells (MFCs have the potential to simultaneously treat wastewater for reuse and to generate electricity. This study mainly considers the performance of an upflow dual-chambered MFC continuously fueled with actual domestic wastewater and alternatively biocatalyzed with aerobic activated sludge and strain of Bacillus Subtilis. The behavior of MFCs during initial biofilm growth and characterization of anodic biofilm were studied. After 45 days of continuous operation, the biofilms on the anodic electrode were well developed. The performance of MFCs was mainly evaluated in terms of COD reductions and electrical power output. Results revealed that the COD removal efficiency was 84% and 90% and the stabilized power outputs were clearly observed achieving a maximum value of 120 and 270 mW/m2 obtained for MFCs inoculated with mixed cultures and Bacillus Subtilis strain, respectively.

  1. Non destructive testing of PWR type fuel pencils by Foucault currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    Marchand, L.

    1984-09-01

    The device presented in this paper, has been developed to satisfy, first, the requirements of nondestructive examinations of irradiated fuel pencils. The technics used allow the acquisition of reliable and detailed information on the state of cans. The computerized data exploitation allows to integrate quickly any improvement factor while allowing to carry out the researches on the automatic diagnostic of Eddy current signals. Now, the equipments of the bench allow visual, metrological and Foucault current simultaneous examinations [fr

  2. Development of Criteria for Flashback Propensity in Jet Flames for High Hydrogen Content and Natural Gas Type Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kalantari, Alireza [Univ. of California, Irvine, CA (United States); Sullivan-Lewis, Elliot [Univ. of California, Irvine, CA (United States); McDonell, Vincent [Univ. of California, Irvine, CA (United States)

    2016-10-17

    Due to increasingly stringent air quality requirements stationary power gas turbines have moved to lean-premixed operation, which reduces pollutant emissions but can result in flashback. Curtailing flashback can be difficult with hydrocarbon fuels and becomes even more challenging when hydrogen is used as the fuel. In fact, flashback is a key operability issue associated with low emission combustion of high hydrogen content fuels. Flashback can cause serious damage to the premixer hardware. Hence, design tools to predict flashback propensity are of interest. Such a design tool has been developed based on the data gathered by experimental study to predict boundary layer flashback using non-dimensional parameters. The flashback propensity of a premixed jet flame has been studied experimentally. Boundary layer flashback has been investigated under turbulent flow conditions at elevated pressures and temperatures (i.e. 3 atm to 8 atm and 300 K to 500 K). The data presented in this study are for hydrogen fuel at various Reynolds numbers, which are representative of practical gas turbine premixer conditions and are significantly higher than results currently available in the literature. Three burner heads constructed of different materials (stainless steel, copper, and zirconia ceramic) were used to evaluate the effect of tip temperature, a parameter found previously to be an important factor in triggering flashback. This study characterizes flashback systematically by developing a comprehensive non-dimensional model which takes into account all effective parameters in boundary layer flashback propensity. The model was optimized for new data and captures the behavior of the new results well. Further, comparison of the model with the single existing study of high pressure jet flame flashback also indicates good agreement. The model developed using the high pressure test rig is able to predict flashback tendencies for a commercial gas turbine engine and can thus serve as a

  3. Emissions from large-scale medium-speed diesel engines: 2. Influence of fuel type and operating mode

    International Nuclear Information System (INIS)

    Sarvi, Arto; Zevenhoven, Ron; Fogelholm, Carl-Johan

    2008-01-01

    This paper addresses gaseous emissions smoke (soot) and particulate matter in large-scale diesel engine exhaust. The test engine was a large-scale turbocharged, after-cooled mean speed (∝ 500 rpm) direct-injection diesel engine and the power per cylinder was about 1 MW. Emission measurements were carried out on burning heavy fuel (HFO) and light fuel (LFO) oils. The test modes for the investigation were a propulsion mode (marine application) and a generator mode (power plant application). Gaseous emissions were measured according to the IMO technical code, smoke (soot) emissions were determined optically and particulate matter (PM) was measured by gravimetric impactor designed for five size fractions. In comparison the emissions from HFO and LFO utilisations indicate slightly higher NO and CO emissions for HFO, while LFO gives clearly higher emissions of hydrocarbons (HC). Emissions of soot and CO appeared to correlate very well, being very high for both fuels throughout the propulsion mode and low load, otherwise being similar for both modes. PM emissions are more than three times higher with HFO than with LFO and appear to decrease with the load except for HFO during the generator mode where an increase of PM emissions with the load is seen. Some data on sampled particles is given. (author)

  4. A prototype machine using thermal type Stirling solar energy and bio fuel as a primary energy source

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Carlos Cesar; Sousa, Regina Celia de; Santos, Jose Maria Ramos dos; Oliveira, Antonio Jose Silva [Universidade Federal do Maranhao (UFMA), Sao Luis, MA (Brazil). Dept. de Fisica

    2011-07-01

    Full text. Depending on the energy crisis and global warming became necessary to seek new sources of energy that could minimize the serious problems arising from this situation. The energy base that supported our growth in recent decades has supported - heavily on fossil fuel, highly polluting since its extraction and consumption, causing great environmental impact. Before his coal, also harmful to human health and nature. Modern life has been moved at the expense of exhaustible resources that took millions of years to form and will end one day. In this work we developed a prototype that uses a heat engine cycle of the Stirling engine with a heat source, arising from the burning of bio fuels or solar power. The main bio fuel used was ethanol. Ethanol is a product of today's diverse market applications, widely used as automotive fuel in hydrated form or blended with gasoline. The main layout of our prototype are: the four-cylinder, two for expansion and the other two for compression, a heat spreader and heat sinks. These simple components can be arranged in various configurations allowing a large space to the adequacy and efficiency of the machine. In experimental measurements made in our prototype, we have an angular speed of 360.1 rpm (revolutions per minute) with an average temperature of 215.6 deg C camera hot (expansion cylinder) and 30 deg C cold source (compression cylinders) and torque generated by our machine is 0.388 Nm Our device is multi-fuel and can be used virtually any source of energy: gasoline, ethanol, methanol, natural gas, diesel, biogas, LPG and solar energy. The construction of this device allowed us to investigate the processes of transformation of energy: chemical, thermal, and mechanical and maximize efficiency of the Stirling engine. To complete the monitoring apparatus, use equipment such as notebook, digital tachometer and a data acquisition Agilent 34970A model. These devices were used in monitoring the angular velocity and

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  6. Bio-based production of fuels and industrial chemicals by repurposing antibiotic-producing type I modular polyketide synthases: opportunities and challenges.

    Science.gov (United States)

    Yuzawa, Satoshi; Keasling, Jay D; Katz, Leonard

    2017-04-01

    Complex polyketides comprise a large number of natural products that have broad application in medicine and agriculture. They are produced in bacteria and fungi from large enzyme complexes named type I modular polyketide synthases (PKSs) that are composed of multifunctional polypeptides containing discrete enzymatic domains organized into modules. The modular nature of PKSs has enabled a multitude of efforts to engineer the PKS genes to produce novel polyketides of predicted structure. We have repurposed PKSs to produce a number of short-chain mono- and di-carboxylic acids and ketones that could have applications as fuels or industrial chemicals.

  7. Experience gathered from the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type

    International Nuclear Information System (INIS)

    Pastorini, A.; Belinco, C.G.; El Bis, E.D.; Sacchi, M.A.; Mayans, C.O.; Martin Ghiselli, A.; Marcora, G.R.

    1990-01-01

    This work describes the needs to materialize the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type, under special conditions, from the Fabrication Pilot Plant sited at the Constituyentes Atomic Center and the Ezeiza Atomic Center, for its subsequent analysis at the High Pressure Experimental Loop. The special conditions under which the transport has been made responded to the fact that the prototype presents a fragile adjustment between rods and separators, necessary to be preserved. (Author) [es

  8. Bio-based production of fuels and industrial chemicals by repurposing antibiotic-producing type I modular polyketide synthases: opportunities and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Yuzawa, Satoshi [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Biological Systems and Engineering Division; Keasling, Jay D. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Biological Systems and Engineering Division; Univ. of California, Berkeley, CA (United States). QB3 Inst.; Joint BioEnergy Inst. (JBEI), Emeryville, CA (United States); Univ. of California, Berkeley, CA (United States). Dept. of Bioengineering; Univ. of California, Berkeley, CA (United States). Dept. of Chemical and Biomolecular Engineering; Technical Univ. of Denmark, Horsholm (Denmark). Novo Nordisk Foundation Center for Biosustainability; Katz, Leonard [Univ. of California, Berkeley, CA (United States). QB3 Inst.

    2016-11-16

    Complex polyketides comprise a large number of natural products that have broad application in medicine and agriculture. They are produced in bacteria and fungi from large enzyme complexes named type I modular polyketide synthases (PKSs) that are composed of multifunctional polypeptides containing discrete enzymatic domains organized into modules. The modular nature of PKSs has enabled a multitude of efforts to engineer the PKS genes to produce novel polyketides of predicted structure. Finally, we have repurposed PKSs to produce a number of short-chain mono- and di-carboxylic acids and ketones that could have applications as fuels or industrial chemicals.

  9. Nondestructive analysis of irradiated fuels

    International Nuclear Information System (INIS)

    Dudey, N.D.; Frick, D.C.

    1977-01-01

    The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

  10. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  11. Influence of fuel type, dilution and equivalence ratio on the emission reduction from the auto-ignition in an Homogeneous Charge Compression Ignition engine

    Energy Technology Data Exchange (ETDEWEB)

    Machrafi, Hatim [UPMC Universite Paris 06, ENSCP, 11 rue de Pierre et Marie Curie, 75005 Paris (France); UPMC Universite Paris 06, Institut Jean Le Rond D' Alembert, 4 place Jussieu, 75252 Paris cedex 05 (France); Universite Libre de Bruxelles, TIPs - Fluid Physics, CP165/67, 50 Avenue F.D. Roosevelt, 1050 Brussels (Belgium); Cavadias, Simeon [UPMC Universite Paris 06, ENSCP, 11 rue de Pierre et Marie Curie, 75005 Paris (France); UPMC Universite Paris 06, Institut Jean Le Rond D' Alembert, 4 place Jussieu, 75252 Paris cedex 05 (France); Amouroux, Jacques [UPMC Universite Paris 06, ENSCP, 11 rue de Pierre et Marie Curie, 75005 Paris (France)

    2010-04-15

    One technology that seems to be promising for automobile pollution reduction is the Homogeneous Charge Compression Ignition (HCCI). This technology still faces auto-ignition and emission-control problems. This paper focuses on the emission problem, since it is incumbent to realize engines that pollute less. For this purpose, this paper presents results concerning the measurement of the emissions of CO, NO{sub x}, CO{sub 2}, O{sub 2} and hydrocarbons. HCCI conditions are used, with equivalence ratios between 0.26 and 0.54, inlet temperatures of 70 C and 120 C and compression ratios of 10.2 and 13.5, with different fuel types: gasoline, gasoline surrogate, diesel, diesel surrogate and mixtures of n-heptane/toluene. The effect of dilution is considered for gasoline, while the effect of the equivalence ratio is considered for all the fuels. No significant amount of NO{sub x} has been measured. It appeared that the CO, O{sub 2} and hydrocarbon emissions were reduced by decreasing the toluene content of the fuel and by decreasing the dilution. The opposite holds for CO{sub 2}. The reduction of the hydrocarbon emission appears to compete with the reduction of the CO{sub 2} emission. Diesel seemed to produce less CO and hydrocarbons than gasoline when auto-ignited. An example of emission reduction control is presented in this paper. (author)

  12. Denatured fuel cycles

    International Nuclear Information System (INIS)

    Till, C.E.

    1979-01-01

    This paper traces the history of the denatured fuel concept and discusses the characteristics of fuel cycles based on the concept. The proliferation resistance of denatured fuel cycles, the reactor types they involve, and the limitations they place on energy generation potential are discussed. The paper concludes with some remarks on the outlook for such cycles

  13. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    Salaices, M.; Salaices, E.; Ovando, R.; Esquivias, J.

    2011-11-01

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  14. O encontro entre o desenvolvimento rural sustentável e a promoção da saúde no Guia Alimentar para a População Brasileira

    OpenAIRE

    Oliveira, Nádia Rosana Fernandes de; Jaime, Patricia Constante

    2016-01-01

    Resumo O Guia Alimentar é documento que aborda os princípios e as recomendações de uma alimentação adequada e saudável para a população brasileira, tendo como propósito apoiar a educação alimentar e nutricional e subsidiar políticas e programas nacionais de alimentação e nutrição no setor de saúde e também em outros setores. O objetivo deste estudo foi identificar a interseção entre a promoção da saúde e o desenvolvimento rural sustentável no Guia Alimentar para a População Brasileira. Realiz...

  15. Fly Ash Formation during Suspension-Firing of Biomass. Effects of Residence Time and Fuel-Type

    DEFF Research Database (Denmark)

    Damø, Anne Juul; Jensen, Peter Arendt; Jappe Frandsen, Flemming

    2017-01-01

    particles were subjected to various analyses, including char burnout level, particle size distribution, elemental composition, and particle morphology and composition. Furthermore, the transient release, i.e. the vaporization of the flame-volatile inorganic elements K, Cl and S, from the burning fuel...... particles to the gas phase, has been quantified by using two different calculation methods. The ash formation mechanisms were found to be quite similar for straw and wood. The degree of conversion (char burn-out level) was generally good at residence times ≥ 1s. The size distribution of the residual fly ash...

  16. Manufacturing and testing of fuel cans with barrier coating for LWR type reactors in USA and Japan

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1988-01-01

    Papers on manufacturing methods for fuel cans of zircalloy with barrier coating of zirconium prepared by pressing an internal tube into external one as well as by pressing of two-layer tubes with further rolling are reviewed. Heat treatment based on creation of the assigned gradient of temperature over tube wall cross section in order to change the structure of a thin layer of the outside surfce when conserving the initial structure of the rest cross section is developed to increase corrosion resistance. Eddy current and ultrasound methods for control of quality and thickness of the barrier layer of zirconium are used

  17. Nuclear fuel string assembly

    International Nuclear Information System (INIS)

    Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

    1976-01-01

    A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

  18. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

    1990-01-01

    The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

  19. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  20. J.S.Bach variações "Goldberg" : um guia para a formação do homem completo

    OpenAIRE

    Helena Jank

    1988-01-01

    Resumo: Dentre muitas formas possíveis de abordagem para esta obra monumental de J. S. Bach, o presente trabalho traz um enfoque eminentemente humanístico, através do qual transparece a intenção do compositor de estruturar, na forma de tema e variações um ?guia para a formação do homem completo? Esta intenção se manifesta na organização geral da obra, com uma clara divisão em três grupos de variações, nos quais são explorados aspectos técnicos, composicionais e de interpretação. Com base na ...

  1. Hydrogen storage by organic chemical hydrides and hydrogen supply to fuel cells with superheated liquid-film-type catalysis

    International Nuclear Information System (INIS)

    Hodoshima, S.; Shono, A.; Sato, K.; Saito, Y.

    2004-01-01

    Organic chemical hydrides, consisting of decalin / naphthalene and tetralin / naphthalene pairs, have been proposed as the storage medium of hydrogen for operating fuel cells in mobile and static modes. The target values in the DOE Hydrogen Plan, U.S., on storage ( 6.5 wt%, 62.0 kg-H 2 / m 3 ) are met with decalin ( 7.3 wt%, 64.8 kg-H 2 / m 3 ). In addition, existing gas stations and tank lorries are available for storage and supply of hydrogen by utilizing the decalin / naphthalene pair, suggesting that decalin is suitable for operating fuel-cell vehicles. Tetralin dehydrogenation proceeds quite rapidly, assuring a predominant power density, though its storage densities ( 3.0 wt%, 28.2 kg-H 2 / m 3 ) are relatively low. Efficient hydrogen supply from decalin or tetralin by heating at 210-280 o C was attained only with the carbon-supported nano-size metal catalysts in the 'superheated liquid-film states' under reactive distillation conditions, where coke formation over the catalyst surface was prevented. The catalyst layer superheated in the liquid-film states gave high reaction rates and conversions, minimizing the evaporation loss under boiling conditions and exergy loss in hydrogen energy systems. (author)

  2. Severe fuel damage experiments performed in the QUENCH facility with 21-rod bundles of LWR-type

    International Nuclear Information System (INIS)

    Sepold, L.; Hering, W.; Schanz, G.; Scholtyssek, W.; Steinbrueck, M.; Stuckert, J.

    2006-01-01

    The objective of the QUENCH experimental program at the Karlsruhe Research Center is to investigate core degradation and the hydrogen source term that results from quenching/flooding an uncovered core, to examine the physical/chemical behavior of overheated fuel elements under different flooding conditions, and to create a data base for model development and improvement of severe fuel damage (SFD) code systems. The large-scale 21-rod bundle experiments conducted in the QUENCH out-of-pile facility are supported by an extensive separate-effects test program, by modeling activities as well as application and improvement of SFD code systems. International cooperations exist with institutions mainly within the European Union but e.g. also with the Russian Academy of Science (IBRAE, Moscow) and the CSARP program of the USNRC. So far, eleven experiments have been performed, two of them with B 4 C absorber material. Experimental parameters were: the temperature at initiation of reflood, the degree of peroxidation, the quench medium, i.e. water or steam, and its injection rate, the influence of a B 4 C absorber rod, the effect of steam-starved conditions before quench, the influence of air oxidation before quench, and boil-off behavior of a water-filled bundle with subsequent quenching. The paper gives an overview of the QUENCH program with its organizational structure, describes the test facility and the test matrix with selected experimental results. (author)

  3. Development of an alternative process for recovery of uranium from rejected plates in the manufacture of MTR type fuel elements

    International Nuclear Information System (INIS)

    Flores Gonzalez, Jocelyn Natalia

    2011-01-01

    This work discusses the recovery of enriched uranium in U 235 , from fuel plates rejected during the fuel elements manufacturing process for the La Reina Nuclear Studies Center, RECH-1, CCHEN. The plates have an aluminum based alloy coating, AISI-SAE 6061, with U 3 Si 2 powder distributed evenly inside and dispersed in an aluminum matrix. The high cost of enriched uranium means that it must be recovered from plates rejected in the production process because of non-compliance with the plate specifications, and also because some of them undergo destructive testing, to measure the aluminum coating's thickness on each side of the plate. The thickness of the uranium nucleus is measured as well and the size of the defects on the ends of the plate such as 'dog bone' and 'fish tail', that is, for the purposes of quality control. The first step in the process is carried out by dissolving the aluminum in a hot solution of NaOH in order to release the uranium silicide powder that is insoluble in the soda. A second step involves dissolving the uranium silicide in a hot HNO 3 solution, followed by washing and filtering, and then extracting the SX and analyzing its behavior during this stage. During the process 98.9% of the uranium is recovered together with a solution that is enough for the SX process given the experiences that were carried out in the extraction stage

  4. Fuel performance, design and development

    International Nuclear Information System (INIS)

    Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

    2006-01-01

    The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Kawai, Mitsuo.

    1988-01-01

    Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

  6. The Bacterial Flagellar Type III Export Gate Complex Is a Dual Fuel Engine That Can Use Both H+ and Na+ for Flagellar Protein Export.

    Directory of Open Access Journals (Sweden)

    Tohru Minamino

    2016-03-01

    Full Text Available The bacterial flagellar type III export apparatus utilizes ATP and proton motive force (PMF to transport flagellar proteins to the distal end of the growing flagellar structure for self-assembly. The transmembrane export gate complex is a H+-protein antiporter, of which activity is greatly augmented by an associated cytoplasmic ATPase complex. Here, we report that the export gate complex can use sodium motive force (SMF in addition to PMF across the cytoplasmic membrane to drive protein export. Protein export was considerably reduced in the absence of the ATPase complex and a pH gradient across the membrane, but Na+ increased it dramatically. Phenamil, a blocker of Na+ translocation, inhibited protein export. Overexpression of FlhA increased the intracellular Na+ concentration in the presence of 100 mM NaCl but not in its absence, suggesting that FlhA acts as a Na+ channel. In wild-type cells, however, neither Na+ nor phenamil affected protein export, indicating that the Na+ channel activity of FlhA is suppressed by the ATPase complex. We propose that the export gate by itself is a dual fuel engine that uses both PMF and SMF for protein export and that the ATPase complex switches this dual fuel engine into a PMF-driven export machinery to become much more robust against environmental changes in external pH and Na+ concentration.

  7. The Bacterial Flagellar Type III Export Gate Complex Is a Dual Fuel Engine That Can Use Both H+ and Na+ for Flagellar Protein Export.

    Science.gov (United States)

    Minamino, Tohru; Morimoto, Yusuke V; Hara, Noritaka; Aldridge, Phillip D; Namba, Keiichi

    2016-03-01

    The bacterial flagellar type III export apparatus utilizes ATP and proton motive force (PMF) to transport flagellar proteins to the distal end of the growing flagellar structure for self-assembly. The transmembrane export gate complex is a H+-protein antiporter, of which activity is greatly augmented by an associated cytoplasmic ATPase complex. Here, we report that the export gate complex can use sodium motive force (SMF) in addition to PMF across the cytoplasmic membrane to drive protein export. Protein export was considerably reduced in the absence of the ATPase complex and a pH gradient across the membrane, but Na+ increased it dramatically. Phenamil, a blocker of Na+ translocation, inhibited protein export. Overexpression of FlhA increased the intracellular Na+ concentration in the presence of 100 mM NaCl but not in its absence, suggesting that FlhA acts as a Na+ channel. In wild-type cells, however, neither Na+ nor phenamil affected protein export, indicating that the Na+ channel activity of FlhA is suppressed by the ATPase complex. We propose that the export gate by itself is a dual fuel engine that uses both PMF and SMF for protein export and that the ATPase complex switches this dual fuel engine into a PMF-driven export machinery to become much more robust against environmental changes in external pH and Na+ concentration.

  8. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  9. INTER-RELAÇÃO DA FERRAMENTA APQP E DO GUIA PMBOK PARA EFICIÊNCIA NAS ETAPAS DE IMPLEMENTAÇÃO DE NOVOS PROJETOS NA INDÚSTRIA AUTOMOTIVA

    Directory of Open Access Journals (Sweden)

    Jean Cristhiano Franco

    2014-06-01

    Full Text Available O desenvolvimento de novos projetos na industria automotiva apresenta-se como um projeto de grande porte, regido por metodologias, guias e ferramentas estruturadas que visam garantir a eficácia do projeto. Esta pesquisa foi desenvolvida com o intuito de contribuir para um maior entendimento entre a ferramenta APQP, de grande aplicação na industria automotiva, e o guia PMBOK, considerado referência no gerenciamento de projetos. Neste âmbito, este estudo realiza uma análise comparativa entre suas principais características, identificando suas semelhanças e divergências. Dentre os diversos resultados do estudo, é possível identificar que o guia PMBOK é mais completo que a ferramenta APQP utilizada no setor automobilístico, uma vez que traz de forma mais detalhada as etapas e as técnicas que são indicadas nas atividades, segundo as boas práticas dos PMPs (Project Management Professional.

  10. Chemical Looping Combustion with Different Types of Liquid Fuels Combustion en boucle chimique avec différentes charges liquides

    Directory of Open Access Journals (Sweden)

    Hoteit A.

    2011-02-01

    Full Text Available CLC is a new promising combustion process for CO2 capture with less or even no energy penalty compared to other processes. Up to now, most of the work performed on CLC was conducted with gaseous or solid fuels, using methane and coal and/or pet coke. Liquid fuels such as heavy fuels resulting from oil distillation or conversion may also be interesting feedstocks to consider. However, liquid fuels are challenging feedstock to deal with in fluidized beds. The objective of the present work is therefore to investigate the feasibility of liquid feed injection and contact with oxygen carrier in CLC conditions in order to conduct partial or complete combustion of hydrocarbons. A batch experimental fluidized bed set-up was developed to contact alternatively oxygen carrier with liquid fuels or air. The 20 mm i.d. fluidized bed reactor was filled up with 45 g of NiAl0.44O1.67 and pulses of 1-2 g of liquid were injected in the bed at high temperatures up to 950˚C. Different feedstocks have been injected, from dodecane to heavy fuel oils No.2. Results show that, during the reduction period, it is possible to convert all the fuel injected and there is no coke remaining on particles at the end of the reduction step. Depending upon oxygen available in the bed, either full combustion or partial combustion can be achieved. Similar results were found with different liquid feeds, despite their different composition and properties. Le CLC est un nouveau concept prometteur appliqué à la combustion qui permet le captage de CO en minimisant la pénalité énergétique liée au captage. Jusqu’à présent, l’essentiel des travaux de recherche dans le domaine du CLC concerne les charges gazeuses (méthane et solides (charbon et coke. Les charges liquides, et particulièrement les résidus pétroliers, sont des charges également intéressantes à considérer a priori. La mise en oeuvre de ces charges en lit fluidisé est cependant délicate. L’objet de ce

  11. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

    1997-01-01

    In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

  13. Guia para estudiantes: Ayuda economica, 2002-2003 (The Student Guide: Financial Aid, 2002-2003).

    Science.gov (United States)

    Department of Education, Washington, DC. Student Financial Assistance.

    This Spanish-language publication explains what federal student financial aid is and what types of student aid are available. The introductory section, "Student Aid at a Glance," presents information about what student aid is, who gets it, and how to get it. The second section discusses "Finding out about Student Aid." The next…

  14. Fuels characterization studies. [jet fuels

    Science.gov (United States)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  15. Supercritical fluid chromatography and high temperature liquid chromatography for the group-type separation of diesel fuels and heavy gas oils

    Energy Technology Data Exchange (ETDEWEB)

    Paproski, R.E.

    2008-07-01

    This thesis investigated the use of unconventional extraction columns for separating diesel fuels by supercritical fluid chromatography (SFC) and for separating heavy gas oils by high temperature normal phase high performance liquid chromatography (HPLC). The purpose was to improve group-type resolution of the fuels, although these methods are also commonly used to determine the proportion of saturates, mono-, di-, tri-, and polyaromatic hydrocarbons. Higher mobile phase flow rates and unconventional column dimensions were also studied to obtain faster analysis times with both SFC and HPLC. The highest group-type resolutions with SFC were obtained by serially coupling bare titania and bare silica columns. Short packed columns and monolithic silica columns were compared at high carbon dioxide flow rates for reducing SFC analysis time, with shortpacked columns achieving 7-fold lower separation times while maintaining significant resolution. Three diesel samples had better resolution and analysis time. A thermally stable bare zircoma column for normal phase HPLC was studied at temperatures up to 200 degrees C. An increase in temperature resulted in lower retention of twenty five aromatic model compounds. Considerable improvements in peak shape, efficiency, group-type selectivity, and column re-equilibration times were obtained at elevated temperatures. At temperatures over 100 degrees C, indole and carbazole thermally decomposed in a hexane/dichloromethane mobile phase. The first order decomposition of carbazole was studied in further detail. A high resolution method was developed using titania and silica columns with valve-switching and dual gradients to separate 3 heavy gas oils. Separation was achieved in only 3 minutes using a fast analysis time method in a titania column at high flow rates.

  16. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Glazoff, Michael V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Eiden, Thomas J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Rezvoi, Aleksey V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed

  17. Seventh Edition Fuel Cell Handbook

    Energy Technology Data Exchange (ETDEWEB)

    NETL

    2004-11-01

    Provides an overview of fuel cell technology and research projects. Discusses the basic workings of fuel cells and their system components, main fuel cell types, their characteristics, and their development status, as well as a discussion of potential fuel cell applications.

  18. Alteration mechanisms of UOX spent fuel under water; Mecanismes d'alteration sous eau du combustible irradie de type UOX

    Energy Technology Data Exchange (ETDEWEB)

    Muzeau, B

    2008-06-15

    The mechanisms of spent fuel alteration in aqueous media need to be understood on the assumption of a direct disposal of the assemblies in a geological formation or for long duration storage in pool. This work is a contribution to the study of the effects of the alpha and/or beta/gamma radiolysis of water on the oxidation and the dissolution of the UO{sub 2} matrix of UOX spent fuel. The effects of the alpha radiolysis, predominant in geological disposal conditions, were quantified by using samples of UO{sub 2} doped with plutonium. The leaching experiments highlighted two types of control for the matrix alteration according to the alpha activity. The first is based on the radiolytic oxidation of the surface and leads to a continuous release of uranium in solution whereas the second is based on a control by the solubility of uranium. An activity threshold, between 18 MBq.g{sup -1} and 33 MBq.g{sup -1}, was defined in a carbonated water. The value of this threshold is dependent on the experimental conditions and the presence or not of electro-active species such as hydrogen in the system. The effects of the alpha/beta/gamma radiolysis in relation with the storage conditions were also quantified. The experimental data obtained on spent fuel indicate that the alteration rate of the matrix based on the behaviour of tracer elements (caesium and strontium) reached a maximum value of some mg.m{sup -2}.d{sup -1}, even under very oxidizing conditions. The solubility of uranium and the nature of the secondary phases depend however on the extent of the oxidizing conditions. (author)

  19. Alteration mechanisms of UOX spent fuel in aqueous media; Mecanismes d'alteration sous eau du combustible irradie de type UOX

    Energy Technology Data Exchange (ETDEWEB)

    Muzeau, B

    2007-06-15

    The mechanisms of underwater alteration of spent fuels need to be understood on the assumption of a direct disposal of the assemblies in a geological formation or for long duration storage in pool. This work is a contribution to the study of the effects of the alpha and/or beta/gamma radiolysis of water on the oxidation and the dissolution of the UO{sub 2} matrix of UOX spent fuel. The effects of the alpha radiolysis, predominant in geological disposal conditions, were quantified using samples of UO{sub 2} doped with plutonium. The leaching experiments highlighted two types of control for the matrix alteration according to the alpha activity. The first is based on the radiolytic oxidation of the surface and leads to a continuous release of uranium in solution whereas the second is based on a control by the solubility of uranium. An activity threshold, located between 18 MBq/g and 33 MBq/g, was defined in a carbonated water. The value of this threshold is dependent on the experimental conditions and the presence or not of electro-active species such as hydrogen in the system. The effects of the alpha/beta/gamma radiolysis in relation with the storage conditions were also quantified. The experimental data obtained on spent fuel indicate that the alteration rate of the matrix based on the behaviour of tracer elements (caesium and strontium) reached a maximum value of some mg.m{sup -2}.d{sup -1}, even under very oxidizing conditions. The solubility of uranium and the nature of the secondary phases depend however on the extent of the oxidizing conditions. (author